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Sample records for fgr evaluation code

  1. Primary structure of the human fgr proto-oncogene product p55/sup c-fgr/

    Energy Technology Data Exchange (ETDEWEB)

    Katamine, S.; Notario, V.; Rao, C.D.; Miki, T.; Cheah, M.S.C.; Tronick, S.R.; Robbins, K.C.

    1988-01-01

    Normal human c-fgr cDNA clones were constructed by using normal peripheral blood mononuclear cell mRNA as a template. Nucleotide sequence analysis of two such clones revealed a 1,587-base-pair-long open reading frame which predicted the primary amino acid sequence of the c-fgr translational product. Homology of this protein with the v-fgr translational product stretched from codons 128 to 516, where 32 differences among 388 codons were observed. Sequence similarity with human c-src, c-yes, and fyn translations products began at amino acid position 76 of the predicted c-fgr protein and extended nearly to its C-terminus. In contrast, the stretch of 75 amino acids at the N-terminus demonstrated a greatly reduced degree of relatedness to these same proteins. To verify the deduced amino acid sequence, antibodies were prepared against peptides representing amino- and carboxy-terminal regions of the predicted c-fgr translational product. Both antibodies specifically recognized a 55-kilodalton protein expressed in COS-1 cells transfected with a c-fgr cDNA expression plasmid. Moreover, the same protein was immunoprecipitated from an Epstein-Barr virus-infected Burkitt's lymphoma cell line which expressed c-fgr mRNA but not in its uninfected fgr mRNA-negative counterpart. These findings identified the 55-kilodalton protein as the product of the human fgr proto-oncogene.

  2. 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

  3. Relationship Between Short Term Variability (STV and Onset of Cerebral Hemorrhage at Ischemia–Reperfusion Load in Fetal Growth Restricted (FGR Mice

    Directory of Open Access Journals (Sweden)

    Takahiro Minato

    2018-05-01

    Full Text Available Fetal growth restriction (FGR is a risk factor exacerbating a poor neurological prognosis at birth. A disease exacerbating a poor neurological prognosis is cerebral palsy. One of the cause of this disease is cerebral hemorrhage including intraventricular hemorrhage. It is believed to be caused by an inability to autoregulate cerebral blood flow as well as immaturity of cerebral vessels. Therefore, if we can evaluate the function of autonomic nerve, cerebral hemorrhage risk can be predicted beforehand and appropriate delivery management may be possible. Here dysfunction of autonomic nerve in mouse FGR fetuses was evaluated and the relationship with cerebral hemorrhage incidence when applying hypoxic load to resemble the brain condition at the time of delivery was examined. Furthermore, FGR incidence on cerebral nerve development and differentiation was examined at the gene expression level. FGR model fetuses were prepared by ligating uterine arteries to reduce placental blood flow. To compare autonomic nerve function in FGR mice with that in control mice, fetal short term variability (STV was measured from electrocardiograms. In the FGR group, a significant decrease in the STV was observed and dysfunction of cardiac autonomic control was confirmed. Among genes related to nerve development and differentiation, Ntrk and Neuregulin 1, which are necessary for neural differentiation and plasticity, were expressed at reduced levels in FGR fetuses. Under normal conditions, Neurogenin 1 and Neurogenin 2 are expressed mid-embryogenesis and are related to neural differentiation, but they are not expressed during late embryonic development. The expression of these two genes increased in FGR fetuses, suggesting that neural differentiation is delayed with FGR. Uterine and ovarian arteries were clipped and periodically opened to give a hypoxic load mimicking the time of labor, and the bleeding rate significantly increased in the FGR group. This suggests that

  4. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

    The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and dose assessment.

  5. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR

    International Nuclear Information System (INIS)

    Kim, So Ra; Min, Byung Il; Park, Kihyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and dose assessment

  6. IFPE/FUMEX-II/CASE27, 7 idealised cases for functional dependence of FGR predictions

    International Nuclear Information System (INIS)

    Turnbull, J.A.; Rossiter, Glyn; Sontheimer, Fritz; Tayal, Mukesh

    2004-01-01

    Description: Seven idealised cases to illustrate the functional dependence of fission gas release (FGR) predictions. (1) Temperature vs Bu for onset of FGR (draft available); (2a) FGR for constant 15 kW/m to 100 MWd/kgU; (2b) FGR for 20 kW/m at BOL decreasing linearly to 10 kW/m at 100 MWd/kgU; (2c) FGR for more realistic power histories supplied by BNFL; (2d) FGR for idealized 'real' histories supplied by FANP; (3a) Candu-Effect of Power on Fission Gas Release; (3b) Candu-Effect of Power Envelope on Fuel Performance

  7. Evaluation the total exposure of soil sample in Adaya site and the obtain risk assessments for the worker by Res Rad code program

    International Nuclear Information System (INIS)

    Mahadi, A. M.; Khadim, A. A. N.; Ibrahim, Z. H.; Ali, S. A.

    2012-12-01

    The present study aims to evaluation the total exposure to the worker in Adaya site risk assessment by using Res Rad code program. The study including 5 areas soil sample calculate in the site and analysis it by High Pure Germaniums (Hg) system made (CANBERRA) company. The soil sample simulation by (Res Rad) code program by inter the radioactive isotope concentration and the specification of the contamination zone area, depth and the cover depth of it. The total exposure of same sample was about 9 mSv/year and the (Heast 2001 Morbidity, FGR13 Morbidity) about 2.045 state every 100 worker in the year. There are simple different between Heast 2001 Morbidity and FGR13 Morbidity according to the Dose Conversion Factor (DCF) use it. The (FGR13 Morbidity) about 2.041 state every 100 worker in the year. (Author)

  8. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  9. 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)

  10. EG-VEGF controls placental growth and survival in normal and pathological pregnancies: case of fetal growth restriction (FGR).

    Science.gov (United States)

    Brouillet, S; Murthi, P; Hoffmann, P; Salomon, A; Sergent, F; De Mazancourt, P; Dakouane-Giudicelli, M; Dieudonné, M N; Rozenberg, P; Vaiman, D; Barbaux, S; Benharouga, M; Feige, J-J; Alfaidy, N

    2013-02-01

    Identifiable causes of fetal growth restriction (FGR) account for 30 % of cases, but the remainders are idiopathic and are frequently associated with placental dysfunction. We have shown that the angiogenic factor endocrine gland-derived VEGF (EG-VEGF) and its receptors, prokineticin receptor 1 (PROKR1) and 2, (1) are abundantly expressed in human placenta, (2) are up-regulated by hypoxia, (3) control trophoblast invasion, and that EG-VEGF circulating levels are the highest during the first trimester of pregnancy, the period of important placental growth. These findings suggest that EG-VEGF/PROKR1 and 2 might be involved in normal and FGR placental development. To test this hypothesis, we used placental explants, primary trophoblast cultures, and placental and serum samples collected from FGR and age-matched control women. Our results show that (1) EG-VEGF increases trophoblast proliferation ([(3)H]-thymidine incorporation and Ki67-staining) via the homeobox-gene, HLX (2) the proliferative effect involves PROKR1 but not PROKR2, (3) EG-VEGF does not affect syncytium formation (measurement of syncytin 1 and 2 and β hCG production) (4) EG-VEGF increases the vascularization of the placental villi and insures their survival, (5) EG-VEGF, PROKR1, and PROKR2 mRNA and protein levels are significantly elevated in FGR placentas, and (6) EG-VEGF circulating levels are significantly higher in FGR patients. Altogether, our results identify EG-VEGF as a new placental growth factor acting during the first trimester of pregnancy, established its mechanism of action, and provide evidence for its deregulation in FGR. We propose that EG-VEGF/PROKR1 and 2 increases occur in FGR as a compensatory mechanism to insure proper pregnancy progress.

  11. A Study on Effect of Recirculated Exhaust Gas upon Performance and Exhaust Emissions in a Power Plant Boiler with FGR System

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Myung-whan; Jung, Kwong-ho; Park, Sung-bum [Gyeongsang Nat’l Univ., Jinju (Korea, Republic of)

    2016-04-15

    The effect of recirculated exhaust gas on performance and exhaust emissions with FGR rate are investigated by using a natural circulation, pressurized draft and water tube boiler with FGR system operating at several boiler loads and over fire air damper openings. The purpose of this study is to apply the FGR system to a power plant boiler for reducing NOx emissions. To activate the combustion, the OFA with 0 to 20% is supplied into the flame. When the suction damper of two stage combustion system installed in the upper side of wind box is opened by handling the lever between 0° and 90°, also, the combustion air supplied to burner is changed. It is found that the fuel consumption rate per evaporation rate did not show an obvious tendency to increase or decrease with rising the FGR rate, and NOx emissions at the same OFA damper opening are decreased, as FGR rates are elevated and boiler loads are dropped. While a trace amount of soot is emitted without regard to the operation conditions of boiler load, OFA damper opening and FGR rate, because soot emissions are eliminated by the electrostatic precipitator with a collecting efficiency of 86.7%.

  12. A Study on Effect of Recirculated Exhaust Gas upon Performance and Exhaust Emissions in a Power Plant Boiler with FGR System

    International Nuclear Information System (INIS)

    Bae, Myung-whan; Jung, Kwong-ho; Park, Sung-bum

    2016-01-01

    The effect of recirculated exhaust gas on performance and exhaust emissions with FGR rate are investigated by using a natural circulation, pressurized draft and water tube boiler with FGR system operating at several boiler loads and over fire air damper openings. The purpose of this study is to apply the FGR system to a power plant boiler for reducing NOx emissions. To activate the combustion, the OFA with 0 to 20% is supplied into the flame. When the suction damper of two stage combustion system installed in the upper side of wind box is opened by handling the lever between 0° and 90°, also, the combustion air supplied to burner is changed. It is found that the fuel consumption rate per evaporation rate did not show an obvious tendency to increase or decrease with rising the FGR rate, and NOx emissions at the same OFA damper opening are decreased, as FGR rates are elevated and boiler loads are dropped. While a trace amount of soot is emitted without regard to the operation conditions of boiler load, OFA damper opening and FGR rate, because soot emissions are eliminated by the electrostatic precipitator with a collecting efficiency of 86.7%.

  13. Validation of fuel performance codes at the NRI Rez plc for Temelin and Dukovany NPPs fuel safety evaluations and operation support

    International Nuclear Information System (INIS)

    Valach, M.; Hejna, J.; Zymak, J.

    2003-05-01

    The report summarises the first phase of the FUMEX II related work performed in the period September 2002 - May 2003. An inventory of the PIN and FRAS codes family used and developed during previous years was made in light of their applicability (validity) in the domain of high burn-up and FUMEX II Project Experimental database. KOLA data were chosen as appropriate for the first step of both codes fixing (both tuned for VVER fuel originally). The modern requirements, expressed by adaptation of the UO 2 conductivity degradation from OECD HRP, RIM and FGR (athermal) modelling implementation into the PIN code and a diffusion FGR model development planned for embedding, into this code allow us to reasonably shadow or keep tight contact with top quality models as TRANSURANUS, COPERNIC, CYRANO, FEMAXI, FRAPCON3 or ENIGMA. Testing and validation runs with prepared input KOLA deck were made. FUMEX II exercise propose LOCA and RIA like transients, so we started development of those two codes coupling - denominated as PIN2FRAS code. Principles of the interface were tested, benchmarking on tentative RIA pulses on highly burned KOLA fuel are presented as the first achievement from our work. (author)

  14. The effect of fuel micro-structure and burn-up on FGR and PCMI studied in IFA-534.13

    International Nuclear Information System (INIS)

    Matsson, I.; Teshima, H.

    1998-02-01

    Fission gas pressure (FGR) and cladding elongation (PCMI) data of four high burnup PWR fuel rods with different grain size (8.5 and 22.1 μm) have been analysed and compared in the IFA-534.13 experiment. The fission gas release is low for both fuel types. During the first part of the irradiation there is no significant difference between the normal grain size fuel and the large grain size fuel. During the second part of the experiment , the FGR appears to be higher in the large grain size fuel. However, this result should be taken with some reservation since the bellows pressure transducer showed signs of irregular behaviour during this period. The FGR at end-of-life in the large grain size fuel is #approx=#2.1 %. The FGR at end-of-life in the normal grain size fuel is #approx=#1.5 %. The degree of PCMI is higher in the large grain size fuel during the first part of the irradiation. During the second period the difference is very small. The point of interaction for PCMI during power ramps has shifted to lower power between beginning and end of irradiation. The two fuel types exhibit very similar behaviour during power ramps. There is no clear indication of relaxation during the irradiation. (author)

  15. A fuel performance code TRUST VIc and its validation

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1997-08-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

  16. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    Ishida, M.; Kogai, T.

    1997-01-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

  17. Evaluation Codes from an Affine Veriety Code Perspective

    DEFF Research Database (Denmark)

    Geil, Hans Olav

    2008-01-01

    Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... includes a reformulation of the usual methods to estimate the minimum distances of evaluation codes into the setting of affine variety codes. Finally we describe the connection to the theory of one-pointgeometric Goppa codes. Contents 4.1 Introduction...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

  18. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  19. Fuel Rod Performance Evaluation of CE 16 x 16 LTA Operated at Steady State Using Transuranus and Pad Codes

    Energy Technology Data Exchange (ETDEWEB)

    Krasnorutskyy, V.; Slyeptsov, O. [Nuclear Fuel Cycle Science and Technology Establishment (NFCSTE), National Science Center, Kharkhov Institute of Physics and Technology (NSC KIPT), Kharkhov (Ukraine)

    2013-03-15

    The report performed under IAEA research contract No. 15370 describes the results of fuel performance evaluation of PWR fuel rods operated at steady state up to discharge burnup of {approx}60 GWD/MTU using the codes of TRANSURANUS designed by ITU and PAD designed by Westinghouse. The experimental results from US-PWR 16x16 LTA Extended Burnup Demonstration Program presented in the IFPE database of the OECD/NEA have been utilized for assessing the codes themselves during simulation of such properties as rod burnup, cladding corrosion, fuel densification and swelling, cladding irradiation growth and strain, FGR and RIP. The results obtained by PAD showed that the code properly simulates rod burnup, cladding irradiation growth and cladding oxidation with Standard Zr-4 material. The calculated burnup values along the fuel stack vary within {+-} 5% of the rod average burnup. The predicted values of the rod axial growth are (0.88-0.94) % and within the measured ones obtained in the burnup range of (50 - 60) GWD/MTU. With allowance made for probability of crud deposition and hot channel hydraulic diameter variation, the axial distribution of oxide layer is predicted well. For the nominal rod dimensions and operation conditions, the calculated peak oxide thickness is slightly overestimated based on the BE corrosion model parameters. The WEC fuel swelling and densification model together with the US NRC one, which is incorporated in the code, were used to assess the change in fuel pellet density ({Delta}{rho}) and fuel volume ({Delta}V{sub F}/V) vs. burnup as well as the rod void volume change, {Delta}V{sub V}/V, and the cladding outer diameter (OD) variation along the fuel stack. (author)

  20. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  1. Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes

    International Nuclear Information System (INIS)

    Sagrado Garcia, I. C.; Vallejo, I.; Herranz, L. E.

    2013-01-01

    The need of defining new RIA safety criteria has pointed out the importance of performing a rigorous assessment of the transient codes capabilities. The present work is a comparative exercise devoted to identify the origin of the key deviations found between the predictions of FRAPTRAN-1.4 and SCANAIR-7.1. To do so, the calculations submitted by CIEMAT to the OECD/NEA RIA benchmark have been exploited. This work shows that deviations in clad temperatures mainly come from the treatment of the oxide layer. The systematically higher deformations calculated by FRAPTRAN-1.4 in early failed tests are caused by the different gap closure estimation. Besides, the dissimilarities observed in the FGR predictions are inherent to the different modeling strategies adopted in each code.

  2. Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes

    Energy Technology Data Exchange (ETDEWEB)

    Sagrado Garcia, I. C.; Vallejo, I.; Herranz, L. E.

    2013-07-01

    The need of defining new RIA safety criteria has pointed out the importance of performing a rigorous assessment of the transient codes capabilities. The present work is a comparative exercise devoted to identify the origin of the key deviations found between the predictions of FRAPTRAN-1.4 and SCANAIR-7.1. To do so, the calculations submitted by CIEMAT to the OECD/NEA RIA benchmark have been exploited. This work shows that deviations in clad temperatures mainly come from the treatment of the oxide layer. The systematically higher deformations calculated by FRAPTRAN-1.4 in early failed tests are caused by the different gap closure estimation. Besides, the dissimilarities observed in the FGR predictions are inherent to the different modeling strategies adopted in each code.

  3. Evaluation of placenta in foetal demise and foetal growth restriction.

    Science.gov (United States)

    Ch, Ujwala; Guruvare, Shyamala; Bhat, Sudha S; Rai, Lavanya; Rao, Sugandhi

    2013-11-01

    The study objective was to evaluate the pathological changes of the placenta in foetal death and foetal growth restriction and to find correlation of the findings with clinical causes. Prospective study at a tertiary care hospital. Gross and histopathological examinations of the placentae were carried out in pregnancies with foetal demise (IUD) and Foetal Growth Restriction (FGR). SPSS, version 11.5. Placentae of twenty seven women with foetal demise and of equal number of women with foetal growth restriction were studied. Placental weight was less than 10(th) percentile in 61.5% women in IUD group and in 93% women in the FGR group. Gross examination of placentae showed abnormalities in 12 (44%) women of IUD group and in 16 (59%) women of FGR group. Histopathological abnormalities were observed in 74.1% women of the IUD group and in 66.7% women of FGR group. Placental histopathology correlated with clinical risk factors in 60% women of IUD group and in 40% women of FGR group. Among the women with no clinically explainable cause for IUD and FGR, 86% and 57% had placental histopathological abnormalities respectively. The histopathological abnormalities of the placenta can be used to document the clinical causes of foetal demise and growth restriction; it may explain the causes in cases of clinically unexplained foetal demise and foetal growth restriction.

  4. Added value of cerebro-placental ratio and uterine artery Doppler at routine third trimester screening as a predictor of SGA and FGR in non-selected pregnancies.

    Science.gov (United States)

    Rial-Crestelo, M; Martinez-Portilla, R J; Cancemi, A; Caradeux, J; Fernandez, L; Peguero, A; Gratacos, E; Figueras, Francesc

    2018-03-04

    The objective of this study is to determine the added value of cerebroplacental ratio (CPR) and uterine Doppler velocimetry at third trimester scan in an unselected obstetric population to predict smallness and growth restriction. We constructed a prospective cohort study of women with singleton pregnancies attended for routine third trimester screening (32 +0 -34 +6 weeks). Fetal biometry and fetal-maternal Doppler ultrasound examinations were performed by certified sonographers. The CPR was calculated as a ratio of the middle cerebral artery to the umbilical artery pulsatility indices. Both attending professionals and patients were blinded to the results, except in cases of estimated fetal weight < p10. The association between third trimester Doppler parameters and small for gestational age (SGA) (birth weight <10th centile) and fetal growth restriction (FGR) (birth weight below the third centile) was assessed by logistic regression, where the basal comparison was a model comprising maternal characteristics and estimated fetal weight (EFW). A total of 1030 pregnancies were included. The mean gestational age at scan was 33 weeks (SD 0.6). The addition of CPR and uterine Doppler to maternal characteristics plus EFW improved the explained uncertainty of the predicting models for SGA (15 versus 10%, p < .001) and FGR (12 versus 8%, p = .03). However, the addition of CPR and uterine Doppler to maternal characteristics plus EFW only marginally improved the detection rates for SGA (38 versus 34% for a 10% of false positives) and did not change the predictive performance for FGR. The added value of CPR and uterine Doppler at 33 weeks of gestation for detecting defective growth is poor.

  5. An Evaluation of Automated Code Generation with the PetriCode Approach

    DEFF Research Database (Denmark)

    Simonsen, Kent Inge

    2014-01-01

    Automated code generation is an important element of model driven development methodologies. We have previously proposed an approach for code generation based on Coloured Petri Net models annotated with textual pragmatics for the network protocol domain. In this paper, we present and evaluate thr...... important properties of our approach: platform independence, code integratability, and code readability. The evaluation shows that our approach can generate code for a wide range of platforms which is integratable and readable....

  6. Order functions and evaluation codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pellikaan, Ruud; van Lint, Jack

    1997-01-01

    Based on the notion of an order function we construct and determine the parameters of a class of error-correcting evaluation codes. This class includes the one-point algebraic geometry codes as wella s the generalized Reed-Muller codes and the parameters are detremined without using the heavy...... machinery of algebraic geometry....

  7. Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

    Energy Technology Data Exchange (ETDEWEB)

    Barani, T.; Bruschi, E.; Pizzocri, D. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy); Pastore, G. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Van Uffelen, P. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, P.O. Box 2340, 76125 Karlsruhe (Germany); Williamson, R.L. [Fuel Modeling and Simulation Department, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Luzzi, L., E-mail: Lelio.Luzzi@polimi.it [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, I-20156 Milano (Italy)

    2017-04-01

    The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release

  8. Computer codes for neutron data evaluation

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1979-01-01

    Data compilation codes such as NESTOR and REPSTOR, and NDES (Neutron Data Evaluation System) are mainly discussed. NDES is a code for neutron data evaluation using a TSS terminal, TEKTRONIX 4014. Users of NDES can perform plotting of data and calculation with nuclear models under conversational mode. (author)

  9. Strength evaluation code STEP for brittle materials

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Futakawa, Masatoshi.

    1997-12-01

    In a structural design using brittle materials such as graphite and/or ceramics it is necessary to evaluate the strength of component under complex stress condition. The strength of ceramic materials is said to be influenced by the stress distribution. However, in the structural design criteria simplified stress limits had been adopted without taking account of the strength change with the stress distribution. It is, therefore, important to evaluate the strength of component on the basis of the fracture model for brittle material. Consequently, the strength evaluation program, STEP, on a brittle fracture of ceramic materials based on the competing risk theory had been developed. Two different brittle fracture modes, a surface layer fracture mode dominated by surface flaws and an internal fracture mode by internal flaws, are treated in the STEP code in order to evaluate the strength of brittle fracture. The STEP code uses stress calculation results including complex shape of structures analyzed by the generalized FEM stress analysis code, ABAQUS, so as to be possible to evaluate the strength of brittle fracture for the structures having complicate shapes. This code is, therefore, useful to evaluate the structural integrity of arbitrary shapes of components such as core graphite components in the HTTR, heat exchanger components made of ceramics materials etc. This paper describes the basic equations applying to the STEP code, code system with a combination of the STEP and the ABAQUS codes and the result of the verification analysis. (author)

  10. Analysis of recent fuel-disruption experiments

    International Nuclear Information System (INIS)

    Kramer, J.M.; Kraft, T.E.; DiMelfi, R.J.; Fenske, G.R.; Gruber, E.E.

    1982-01-01

    Recent USDOE-sponsored DEH, FGR, and TREAT F series fuel-disruption experiments are analyzed with existing analytical models. The experiments are interpreted and the results used to evaluate the models. Calculations are presented using the FRAS3 fission-gas-behavior code and the DiMelfi-Deitrich fuel-response model

  11. An analysis of recent fuel disruption experiments

    International Nuclear Information System (INIS)

    Kramer, J.M.; Kraft, T.E.; Dimelfi, R.J.; Fenske, G.R.; Gruber, E.E.

    1982-01-01

    Recent USDOE-Sponsored DEH, FGR, and TREAT F series fuel disruption experiments are analyzed with existing analytical models. The experiments are interpreted and the results used to evaluate the models. Calculations are presented using the FRAS3 fission gas behavior code and the DiMelfi-Deitrich fuel response model

  12. Performance evaluation based on data from code reviews

    OpenAIRE

    Andrej, Sekáč

    2016-01-01

    Context. Modern code review tools such as Gerrit have made available great amounts of code review data from different open source projects as well as other commercial projects. Code reviews are used to keep the quality of produced source code under control but the stored data could also be used for evaluation of the software development process. Objectives. This thesis uses machine learning methods for an approximation of review expert’s performance evaluation function. Due to limitations in ...

  13. Rapport: Coding Class - Dokumentation og evaluering

    DEFF Research Database (Denmark)

    Hansbøl, Mikala; Ejsing-Duun, Stine

    2017-01-01

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet . Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates . Rapporten er forfattet af Docent i digitale læringsressourcer og faglig leder af forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign, design...... tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for Kommunikation ved Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  14. Allele coding in genomic evaluation

    Directory of Open Access Journals (Sweden)

    Christensen Ole F

    2011-06-01

    Full Text Available Abstract Background Genomic data are used in animal breeding to assist genetic evaluation. Several models to estimate genomic breeding values have been studied. In general, two approaches have been used. One approach estimates the marker effects first and then, genomic breeding values are obtained by summing marker effects. In the second approach, genomic breeding values are estimated directly using an equivalent model with a genomic relationship matrix. Allele coding is the method chosen to assign values to the regression coefficients in the statistical model. A common allele coding is zero for the homozygous genotype of the first allele, one for the heterozygote, and two for the homozygous genotype for the other allele. Another common allele coding changes these regression coefficients by subtracting a value from each marker such that the mean of regression coefficients is zero within each marker. We call this centered allele coding. This study considered effects of different allele coding methods on inference. Both marker-based and equivalent models were considered, and restricted maximum likelihood and Bayesian methods were used in inference. Results Theoretical derivations showed that parameter estimates and estimated marker effects in marker-based models are the same irrespective of the allele coding, provided that the model has a fixed general mean. For the equivalent models, the same results hold, even though different allele coding methods lead to different genomic relationship matrices. Calculated genomic breeding values are independent of allele coding when the estimate of the general mean is included into the values. Reliabilities of estimated genomic breeding values calculated using elements of the inverse of the coefficient matrix depend on the allele coding because different allele coding methods imply different models. Finally, allele coding affects the mixing of Markov chain Monte Carlo algorithms, with the centered coding being

  15. Evaluation of Code Blue Implementation Outcomes

    Directory of Open Access Journals (Sweden)

    Bengü Özütürk

    2015-09-01

    Full Text Available Aim: In this study, we aimed to emphasize the importance of Code Blue implementation and to determine deficiencies in this regard. Methods: After obtaining the ethics committee approval, 225 patient’s code blue call data between 2012 and 2014 January were retrospectively analyzed. Age and gender of the patients, date and time of the call and the clinics giving Code Blue, the time needed for the Code Blue team to arrive, the rates of false Code Blue calls, reasons for Code Blue calls and patient outcomes were investigated. Results: A total of 225 patients (149 male, 76 female were evaluated in the study. The mean age of the patients was 54.1 years. 142 (67.2% Code Blue calls occurred after hours and by emergency unit. The mean time for the Code Blue team to arrive was 1.10 minutes. Spontaneous circulation was provided in 137 patients (60.8%; 88 (39.1% died. The most commonly identified possible causes were of cardiac origin. Conclusion: This study showed that Code Blue implementation with a professional team within an efficient and targeted time increase the survival rate. Therefore, we conclude that the application of Code Blue carried out by a trained team is an essential standard in hospitals. (The Medical Bulletin of Haseki 2015; 53:204-8

  16. Evaluation of three coding schemes designed for improved data communication

    Science.gov (United States)

    Snelsire, R. W.

    1974-01-01

    Three coding schemes designed for improved data communication are evaluated. Four block codes are evaluated relative to a quality function, which is a function of both the amount of data rejected and the error rate. The Viterbi maximum likelihood decoding algorithm as a decoding procedure is reviewed. This evaluation is obtained by simulating the system on a digital computer. Short constraint length rate 1/2 quick-look codes are studied, and their performance is compared to general nonsystematic codes.

  17. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    International Nuclear Information System (INIS)

    Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

    1997-01-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

  18. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

    1997-08-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

  19. Data evaluation and code comparison activities

    International Nuclear Information System (INIS)

    Itikawa, Yukikazu; Takagi, Hidekazu; Nakamura, Yoshiharu; Imai, Makoto; Sasaki, Akira

    2013-01-01

    In atomic and molecular data base, intolerable numerical differences beyond error margin are found among some papers resulted from measurements or calculations even for the same collision processes. These differences spoil the reliability of the data base. This report describes the data evaluation for atomic and molecular data promoted by IAEA cooperated with other institutes, which Japanese researchers collaborate with. The reaction cross sections calculated numerically are evaluated for the collisions between electrons and molecular ions of H 2 + and HeH + . The application of an electron swarm parameter was shown for the evaluation and determination of the collision cross sections between electrons and molecules. In order to complete higher precision of atomic codes and a collisional-radiative model, IAEA held the workshop for the code comparison of the nonlocal thermodynamic equilibrium. (Y. Kazumata)

  20. Methodology for Evaluating Cost-effectiveness of Commercial Energy Code Changes

    Energy Technology Data Exchange (ETDEWEB)

    Hart, Philip R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-01-31

    This document lays out the U.S. Department of Energy’s (DOE’s) method for evaluating the cost-effectiveness of energy code proposals and editions. The evaluation is applied to provisions or editions of the American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE) Standard 90.1 and the International Energy Conservation Code (IECC). The method follows standard life-cycle cost (LCC) economic analysis procedures. Cost-effectiveness evaluation requires three steps: 1) evaluating the energy and energy cost savings of code changes, 2) evaluating the incremental and replacement costs related to the changes, and 3) determining the cost-effectiveness of energy code changes based on those costs and savings over time.

  1. Adaptive RAC codes employing statistical channel evaluation ...

    African Journals Online (AJOL)

    An adaptive encoding technique using row and column array (RAC) codes employing a different number of parity columns that depends on the channel state is proposed in this paper. The trellises of the proposed adaptive codes and a statistical channel evaluation technique employing these trellises are designed and ...

  2. Allele coding in genomic evaluation

    DEFF Research Database (Denmark)

    Standen, Ismo; Christensen, Ole Fredslund

    2011-01-01

    Genomic data are used in animal breeding to assist genetic evaluation. Several models to estimate genomic breeding values have been studied. In general, two approaches have been used. One approach estimates the marker effects first and then, genomic breeding values are obtained by summing marker...... effects. In the second approach, genomic breeding values are estimated directly using an equivalent model with a genomic relationship matrix. Allele coding is the method chosen to assign values to the regression coefficients in the statistical model. A common allele coding is zero for the homozygous...... genotype of the first allele, one for the heterozygote, and two for the homozygous genotype for the other allele. Another common allele coding changes these regression coefficients by subtracting a value from each marker such that the mean of regression coefficients is zero within each marker. We call...

  3. Evaluation of the SCANAIR Computer Code

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali

    2001-11-01

    The SCANAIR computer code, version 3.2, has been evaluated from the standpoint of its capability to analyze, simulate and predict nuclear fuel behavior during severe power transients. SCANAIR calculates the thermal and mechanical behavior of a pressurized water reactor (PWR) fuel rod during a postulated reactivity initiated accident (RIA), and our evaluation indicates that SCANAIR is a state of the art computational tool for this purpose. Our evaluation starts by reviewing the basic theoretical models in SCANAIR, namely the governing equations for heat transfer, the mechanical response of fuel and clad, and the fission gas release behavior. The numerical methods used to solve the governing equations are briefly reviewed, and the range of applicability of the models and their limitations are discussed and illustrated with examples. Next, the main features of the SCANAIR user interface are delineated. The code requires an extensive amount of input data, in order to define burnup-dependent initial conditions to the simulated RIA. These data must be provided in a special format by a thermal-mechanical fuel rod analysis code. The user also has to supply the transient power history under RIA as input, which requires a code for neutronics calculation. The programming structure and documentation of the code are also addressed in our evaluation. SCANAIR is programmed in Fortran-77, and makes use of several general Fortran-77 libraries for handling input/output, data storage and graphical presentation of computed results. The documentation of SCANAIR and its helping libraries is generally of good quality. A drawback with SCANAIR in its present form, is that the code and its pre- and post-processors are tied to computers running the Unix or Linux operating systems. As part of our evaluation, we have performed a large number of computations with SCANAIR, some of which are documented in this report. The computations presented here include a hypothetical RIA in a high

  4. Development of Evaluation Code for MUF Uncertainty

    International Nuclear Information System (INIS)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan

    2015-01-01

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities

  5. Development of Evaluation Code for MUF Uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities.

  6. Dante-unfolding code for energy spectra evaluation

    International Nuclear Information System (INIS)

    Petilli, M.

    1979-01-01

    The code DANTE, using the last square method in unfolding for dosimetry purpose, solves the neutron spectra evaluation problem starting by activity measurements. The code DANTE introduced for the first time the correlation between available data by mean of flux and activity variance-covariance matrices and the error propagation. In the present report the solution method is detailed described

  7. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  8. Probabilistic evaluations for CANTUP computer code analysis improvement

    International Nuclear Information System (INIS)

    Florea, S.; Pavelescu, M.

    2004-01-01

    Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

  9. Comparative calculations and operation-to-PIE data juxtaposition of the Zaporozhye NPP, WWER-1000 FA-E0325 fuel rods after 4 years of operation up to ∼49 MWd/kgU burnup

    International Nuclear Information System (INIS)

    Passage, G.; Stefanova, S.; Scheglov, A.; Proselkov, V.

    2006-01-01

    Operational and PIE data for the Zaporozhe NPP, FA-E0325, WWER-1000 fuel rods were provided in the OECD NEA IFPE Database and were used to perform comparative calculations among several fuel performance codes. The fuel rods had been irradiated for 4 years of operation up to ∼49 MWd/kg U burnup. The fuel rod operation histories are developed for the PINw99, TRANSURANUS (V1M1J03) and TOPRA-2 codes. The initial state fuel rod parameters are analysed and calculations are carried out. The PIE data enable the comparison of experimental measurement with code-calculated values for cladding elongation (49 rods), FGR and gas pressure (35 rods). Cladding diameter creep-down and gap closure results are juxtaposed as well. The capability of the applied codes correctly to predict the WWER fuel rod performance is shown. The WWER-1000 fuel rod data include initial geometrical and design parameters of the fuel rods, as well as description of the operation regime, NPP unit loading history and PIE results at normal conditions. The data are sufficient for modelling all 312 fuel rod and for comparison of calculations with experimental results for a limited number of fuel rods. The comparison between the calculated and measured results discussed in this paper shows that the codes PINw99, TRANSURANUS and TOPRA-2, are capable of adequate predicting the thermophysical and the mechanical performance of the WWER-1000 fuel rods. The PINw99 code predicts conservative BOL FGR values and conservative gas pressure values in the region of burnups higher than 30 MWd/kg U, which can be explained by the underprediction of the cladding gas inner volume and cladding elongation. The improved version PIN2K (not applied in the present study) predicts much better FGR and gas pressure, though, it is still under development in the high burnup FGR modelling part. In the TRANSURANUS code, there are also areas, where refinements are clearly indicated. They are subjects of the ongoing research projects and

  10. Levels of neopterin and C-reactive protein in pregnant women with fetal growth restriction.

    Science.gov (United States)

    Erkenekli, K; Keskin, U; Uysal, B; Kurt, Y G; Sadir, S; Çayci, T; Ergün, A; Erkaya, S; Danişman, N; Uygur, D

    2015-04-01

    The aim of this study was to evaluate whether pregnant women with fetal growth restriction (FGR) have higher plasma neopterin and C-reactive protein (CRP) concentrations compared with those with uncomplicated pregnancy. A total of 34 pregnant women with FGR and 62 patients with uncomplicated pregnancy were included. Neopterin and CRP levels were measured at the time of diagnosis. The primary outcome of this study was to compare the neopterin and CRP levels in pregnant women with FGR and those with uncomplicated pregnancies. The secondary outcome of our study was to evaluate the correlation between fetal birth weight and maternal neopterin levels. The serum neopterin levels were significantly elevated in pregnant women with FGR (22.71 ± 7.70 vs 19.15 ± 8.32). However, CRP was not elevated in pregnant women with FGR (7.47 ± 7.59 vs 5.29 ± 3.58). These findings support the hypothesis that pregnancy with FGR is associated with a marked increase in macrophage activation and the natural immune system.

  11. Evaluation of the FRAPCON-3 Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala (Sweden)

    2002-03-01

    The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO{sub 2} fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models

  12. Evaluation of the FRAPCON-3 Computer Code

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali

    2002-03-01

    The FRAPCON-3 computer code has been evaluated with respect to its applicability, modeling capability, user friendliness, source code structure and supporting experimental database. The code is intended for thermo-mechanical analyses of light water reactor nuclear fuel rods under steady-state operational conditions and moderate power excursions. It is applicable to both boiling- and pressurized water reactor fuel rods with UO 2 fuel, ranging up to about 65 MWd/kg U in rod average burnup. The models and numerical methods in FRAPCON-3 are relatively simple, which makes the code transparent and also fairly easy to modify and extend for the user. The fundamental equations for heat transfer, structural analysis and fuel fission gas release are solved in one-dimensional (radial) and stationary (time-independent) form, and interaction between axial segments of the rod is confined to calculations of coolant axial flow and rod internal gas pressure. The code is fairly easy to use; fuel rod design data and time histories of fuel rod power and coolant inlet conditions are input via a single text file, and the corresponding calculated variation with time of important fuel rod parameters are printed to a single output file in textual form. The results can also be presented in graphical form through an interface to the general graphics program xmgr. FRAPCON-3 also provides the possibility to export calculated results to the transient fuel rod analysis code FRAPTRAN, where the data can be used as burnup-dependent initial conditions to a postulated transient. Most of the source code to FRAPCON-3 is written in Fortran-IV, which is an archaic, non-standard dialect of the Fortran programming language. Since Fortran-IV is not accepted by all compilers for the latest standard of the language, Fortran-95, there is a risk that the source code must be partly rewritten in the future. Documentation of the code comprises (i) a general code description, which briefly presents models

  13. Evaluation of practicality of ASME code, Section XI

    International Nuclear Information System (INIS)

    Mattu, R.K.; Lauderdale, J.R.; Liu, S.N.; Lance, J.J.

    2004-01-01

    Many nuclear power plants have found that it is impractical or unduly burdensome to comply with some ASME Boiler and Pressure Code provisions and have sought relief from those provisions from the Nuclear Regulatory Commission. An Electric Power Research Institute (EPRI) project is evaluating such Code provisions and alternatives to them that will meet the safety intent of the Code with less burden on utilities. The methodology is to extract data from an on-line data base of relief requests since 1980, analyse the data to identify burdensome provisions for which there are satisfactory alternatives, and recommend changes in the Code to the ASME. (author)

  14. Extending CANTUP code analysis to probabilistic evaluations

    International Nuclear Information System (INIS)

    Florea, S.

    2001-01-01

    The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

  15. Evaluation of the DRAGON code for VHTR design analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-12

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.

  16. Evaluation of the DRAGON code for VHTR design analysis

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division

    2006-01-01

    This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR

  17. Elastic creep-fatigue evaluation for ASME code

    International Nuclear Information System (INIS)

    Severud, L.K.; Winkel, B.V.

    1987-01-01

    Experience with applying the ASME Code Case N-47 rules for evaluation of creep-fatigue with elastic analysis results has been problematic. The new elastic evaluation methods are intended to bound the stress level and strain range values needed for use in employing the code inelastic analysis creep-fatigue damage counting procedures. To account for elastic followup effects, ad hoc rules for stress classification, shakedown, and ratcheting are employed. Because elastic followup, inelastic strain concentration, and stress-time effects are accounted for, the design fatigue curves in Case N-47 for inelastic analysis are used instead of the more conservative elastic analysis curves. Creep damage assessments are made using an envelope stress-time history that treats multiple load events and repeated cycles during elevated temperature service life. (orig./GL)

  18. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  19. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  20. RF cavity evaluation with the code SUPERFISH

    International Nuclear Information System (INIS)

    Hori, T.; Nakanishi, T.; Ueda, N.

    1982-01-01

    The computer code SUPERFISH calculates axisymmetric rf fields and is most applicable to re-entrant cavities of an Alvarez linac. Some sample results are shown for the first Alvarez's in NUMATRON project. On the other hand the code can also be effectivily applied to TE modes excited in an RFQ linac when the cavity is approximately considered as positioning at an infinite distance from the symmetry axis. The evaluation was made for several RFQ cavities, models I, II and a test linac named LITL, and useful results for the resonator design were obtained. (author)

  1. SALE: Safeguards Analytical Laboratory Evaluation computer code

    International Nuclear Information System (INIS)

    Carroll, D.J.; Bush, W.J.; Dolan, C.A.

    1976-09-01

    The Safeguards Analytical Laboratory Evaluation (SALE) program implements an industry-wide quality control and evaluation system aimed at identifying and reducing analytical chemical measurement errors. Samples of well-characterized materials are distributed to laboratory participants at periodic intervals for determination of uranium or plutonium concentration and isotopic distributions. The results of these determinations are statistically-evaluated, and each participant is informed of the accuracy and precision of his results in a timely manner. The SALE computer code which produces the report is designed to facilitate rapid transmission of this information in order that meaningful quality control will be provided. Various statistical techniques comprise the output of the SALE computer code. Assuming an unbalanced nested design, an analysis of variance is performed in subroutine NEST resulting in a test of significance for time and analyst effects. A trend test is performed in subroutine TREND. Microfilm plots are obtained from subroutine CUMPLT. Within-laboratory standard deviations are calculated in the main program or subroutine VAREST, and between-laboratory standard deviations are calculated in SBLV. Other statistical tests are also performed. Up to 1,500 pieces of data for each nuclear material sampled by 75 (or fewer) laboratories may be analyzed with this code. The input deck necessary to run the program is shown, and input parameters are discussed in detail. Printed output and microfilm plot output are described. Output from a typical SALE run is included as a sample problem

  2. User's guide to the repository intrusion risk evaluation code INTRUDE

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Thorne, M.C.

    1986-05-01

    The report, commissioned by the Department of the Environment as part of its radioactive waste management research programme, constitutes the user's guide to the repository intrusion risk evaluation code INTRUDE. It provides an explanation of the mathematical basis of the code, the database used and the operation of the code. INTRUDE is designed to facilitate the estimation of individual risks arising from the possibility of intrusion into shallow land burial facilities for radioactive wastes. It considers a comprehensive inventory of up to 65 long-lived radionuclides and produces risk estimates for up to 20 modes of intrusion and up to 50 times of evaluation. (author)

  3. Foetal growth restriction is associated with poor reading and spelling skills at eight years to 10 years of age.

    Science.gov (United States)

    Partanen, Lea; Korkalainen, Noora; Mäkikallio, Kaarin; Olsén, Päivi; Laukkanen-Nevala, Päivi; Yliherva, Anneli

    2018-01-01

    Foetal growth restriction (FGR) is associated with communication problems, which might lead to poor literacy skills. The reading and spelling skills of eight- to 10-year-old FGR children born at 24-40 gestational weeks were compared with those of their gestational age-matched, appropriately grown (AGA) peers. A prospectively collected cohort of 37 FGR and 31 AGA children was recruited prenatally at a Finnish tertiary care centre during 1998-2001. The children's reading and spelling skills were assessed using standardised tests for Finnish-speaking second and third graders. Significantly more children performed below the 10th percentile normal values for reading and spelling skills in the FGR group than in the AGA group. At nine years of age, the FGR children had significantly poorer performance in word reading skills and reading fluency, reading accuracy and reading comprehension than the AGA controls. No between-group differences were detected at eight years of age. FGR is associated with poor performance in reading and spelling skills. A third of the FGR children performed below the 10th percentile normal values at nine years of age. These results indicate a need to continuously evaluate linguistic and literacy skills as FGR children age to ensure optimal support. ©2017 Foundation Acta Paediatrica. Published by John Wiley & Sons Ltd.

  4. Strong normalization by type-directed partial evaluation and run-time code generation

    DEFF Research Database (Denmark)

    Balat, Vincent; Danvy, Olivier

    1998-01-01

    We investigate the synergy between type-directed partial evaluation and run-time code generation for the Caml dialect of ML. Type-directed partial evaluation maps simply typed, closed Caml values to a representation of their long βη-normal form. Caml uses a virtual machine and has the capability...... to load byte code at run time. Representing the long βη-normal forms as byte code gives us the ability to strongly normalize higher-order values (i.e., weak head normal forms in ML), to compile the resulting strong normal forms into byte code, and to load this byte code all in one go, at run time. We...... conclude this note with a preview of our current work on scaling up strong normalization by run-time code generation to the Caml module language....

  5. Strong Normalization by Type-Directed Partial Evaluation and Run-Time Code Generation

    DEFF Research Database (Denmark)

    Balat, Vincent; Danvy, Olivier

    1997-01-01

    We investigate the synergy between type-directed partial evaluation and run-time code generation for the Caml dialect of ML. Type-directed partial evaluation maps simply typed, closed Caml values to a representation of their long βη-normal form. Caml uses a virtual machine and has the capability...... to load byte code at run time. Representing the long βη-normal forms as byte code gives us the ability to strongly normalize higher-order values (i.e., weak head normal forms in ML), to compile the resulting strong normal forms into byte code, and to load this byte code all in one go, at run time. We...... conclude this note with a preview of our current work on scaling up strong normalization by run-time code generation to the Caml module language....

  6. Second- to third-trimester longitudinal growth assessment for prediction of small-for-gestational age and late fetal growth restriction.

    Science.gov (United States)

    Caradeux, J; Eixarch, E; Mazarico, E; Basuki, T R; Gratacós, E; Figueras, F

    2018-02-01

    Detection of fetal growth restriction (FGR) remains poor and most screening strategies rely on cross-sectional evaluation of fetal size during the third trimester. A longitudinal and individualized approach has been proposed as an alternative method of evaluation. The aim of this study was to compare second- to third-trimester longitudinal growth assessment to cross-sectional evaluation in the third trimester for the prediction of small-for-gestational age (SGA) and late FGR in low-risk singleton pregnancy. This was a prospective cohort study of 2696 unselected consecutive low-risk singleton pregnancies scanned at 21 ± 2 and 32 ± 2 weeks. For cross-sectional growth assessment, abdominal circumference (AC) measurements were transformed to z-values according the 21st-INTERGROWTH standards. Longitudinal growth assessment was performed by calculating the AC z-velocity and the second- to third-trimester AC conditional growth centile. Longitudinal assessment was compared with cross-sectional assessment at 32 weeks. Association of cross-sectional and longitudinal evaluations with SGA and late FGR was assessed by logistic regression analysis. Predictive performance was determined by receiver-operating characteristics curve analysis. In total, 210 (7.8%) newborns were classified as SGA and 103 (3.8%) as late FGR. Neither longitudinal measurement improved the association with SGA or late FGR provided by cross-sectional evaluation of AC z-score at 32 weeks. Areas under the curves of AC z-velocity and conditional AC growth were significantly smaller than those of cross-sectional AC z-scores (P third trimester has a low predictive capacity for SGA and late FGR in low-risk singleton pregnancy compared with cross-sectional growth evaluation. Copyright © 2017 ISUOG. Published by John Wiley & Sons Ltd. Copyright © 2017 ISUOG. Published by John Wiley & Sons Ltd.

  7. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  8. Evaluation of the analysis models in the ASTRA nuclear design code system

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-11-15

    In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.

  9. Radioactivities evaluation code system for high temperature gas cooled reactors during normal operation

    International Nuclear Information System (INIS)

    Ogura, Kenji; Morimoto, Toshio; Suzuki, Katsuo.

    1979-01-01

    A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)

  10. Modelling Reactivity-Initiated-Accident Experiments With Falcon And SCANAIR: A Comparison Exercise

    International Nuclear Information System (INIS)

    Romano, A.; Wallin, H.; Zimmermann, M.A.

    2005-01-01

    A critical assessment is made of the state-of-the-art fuel performance code FALCON in the context of selected Reactivity Initiated Accident (RIA) experiments from the CABRI REP Na series, and contrasts its predictions against those of the extensively benchmarked SCANAIR (Version 3.2) code. The thermal fields in the fuel and cladding, the clad mechanical deformation, and the Fission Gas Release (FGR) are adopted as 'Figures of Merit' by which to judge code performance. Particular attention is paid to the importance of fission-gas-induced clad deformation (which is modelled in SCANAIR, but not in FALCON), relative to that driven by the fuel thermal expansion (which is modelled by both codes). The thermal fields calculated by the codes are in good agreement with each other, especially during the initial stages of the transients --- the adiabatic phase. Larger discrepancies are observed at later times, and are due to the different models applied to calculate the gap conductance. FALCON predicts clad permanent deformations at the end of the transients with a maximum deviation from the experimental measurements of about 20%. Generally, the code always tends to underpredict the measurements. SCANAIR performs similarly, but grossly overpredicts the permanent clad strain for the case involving a very energetic pulse. The fission-gas-driven clad deformation is only relevant for very fast pulse energy injection cases, which are not prototypical of the RIA transients expected in PWRs. The FGR models in FALCON do not capture the mechanism of 'burst-release' in the RIA transients, having been developed for steady-state irradiation conditions. This also explains why they performed poorly when applied to the fast-transient cases analyzed here. In contrast, the FGR results from SCANAIR are in satisfactory agreement with the experimental results. (author)

  11. A computer code for Tokamak reactor concepts evaluation

    International Nuclear Information System (INIS)

    Rosatelli, F.; Raia, G.

    1985-01-01

    A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

  12. Dosskin code for radiological evaluation of skin radioactive contaminations

    International Nuclear Information System (INIS)

    Cornejo D, N.

    1996-01-01

    The conceptual procedure and computational features of the DOSSKIN code are shown. This code calculates, in a very interactive way, skin equivalent doses and radiological risk related to skin radioactive contaminations. The evaluation takes into account the contributions of contaminant daughter nuclides and backscattering of beta particles in any skin cover. DOSSKIN also allows to estimate the maximum time needed to decontaminate the affected zone, using, as input quantity, the limit value of skin equivalent dose considered by users. The comparison of the results obtained by the DOSSKIN code with those reported by different authors are showed. The differences of results are less than 30%. (authors). 4 refs., 3 fig., 1 tab

  13. Clinical evaluation of coded excitation in medical ultrasound

    DEFF Research Database (Denmark)

    Pedersen, Morten Høgholm; Misaridis, Thanssis; Jensen, Jørgen Arendt

    2003-01-01

    -K Medical model 3535) with transmitter and receiver boards developed in our group and a mechanical 4 MHz transducer were used. The system acquired coded and conventional US image frames interleaved, yielding identical acquisitions with the two techniques. Cine-loop sequences were evaluated by three...

  14. Evaluating Open-Source Full-Text Search Engines for Matching ICD-10 Codes.

    Science.gov (United States)

    Jurcău, Daniel-Alexandru; Stoicu-Tivadar, Vasile

    2016-01-01

    This research presents the results of evaluating multiple free, open-source engines on matching ICD-10 diagnostic codes via full-text searches. The study investigates what it takes to get an accurate match when searching for a specific diagnostic code. For each code the evaluation starts by extracting the words that make up its text and continues with building full-text search queries from the combinations of these words. The queries are then run against all the ICD-10 codes until a match indicates the code in question as a match with the highest relative score. This method identifies the minimum number of words that must be provided in order for the search engines choose the desired entry. The engines analyzed include a popular Java-based full-text search engine, a lightweight engine written in JavaScript which can even execute on the user's browser, and two popular open-source relational database management systems.

  15. Annual report of nuclear code evaluation committee for fiscal 2000 year

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-03-01

    In this report, research results discussed in fiscal 2000 year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. In 2000, papers mainly on the three topics of (1) present status of burnup credit evaluation methods, (2) issues concerning convergence of criticality calculation and (3) estimation methods for errors associated with criticality calculation based on nuclear data covariance file, are presented and discussed. These results are sorted to grasp the present status of related technology and described in this report. (author)

  16. Applicability evaluation on the conservative metal-water reaction(MWR) model implemented into the SPACE code

    International Nuclear Information System (INIS)

    Lee, Suk Ho; You, Sung Chang; Kim, Han Gon

    2011-01-01

    The SBLOCA (Small Break Loss-of-Coolant Accident) evaluation methodology for the APR1400 (Advanced Power Reactor 1400) is under development using the SPACE code. The goal of the development of this methodology is to set up a conservative evaluation methodology in accordance with Appendix K of 10CFR50 by the end of 2012. In order to develop the Appendix K version of the SPACE code, the code modification is considered through implementation of the code on the required evaluation models. For the conservative models required in the SPACE code, the metal-water reaction (MWR) model, the critical flow model, the Critical Heat Flux (CHF) model and the post-CHF model must be implemented in the code. At present, the integration of the model to generate the Appendix K version of SPACE is in its preliminary stage. Among them, the conservative MWR model and its code applicability are introduced in this paper

  17. Development of an inelastic stress analysis code 'KINE-T' and its evaluations

    International Nuclear Information System (INIS)

    Kobatake, K.; Takahashi, S.; Suzuki, M.

    1977-01-01

    Referring to the ASME B and PVC Code Case 1592-7, the inelastic stress analysis is required for the designs of the class 1 components in elevated temperature if the results of the elastic stress analysis and/or simplified inelastic analysis do not satisfy the requirements. Authors programmed a two-dimensional axisymmetric inelastic analysis code 'KINE-T', and carried out its evaluations and an application. This FEM code is based on the incremental method and the following: elastic-plastic constitutive equation (yield condition of von Mises; flow rule of Prandtl-Reuss; Prager's hardening rule); creep constitutive equation (equation of state approach; flow rule of von Mises; strain hardening rule); the temperature dependency of the yield function is considered; solution procedure of the assembled stiffness matrix is the 'initial stress method'. After the completion of the programming, authors compared the output with not only theoretical results but also with those of the MARC code and the ANSYS code. In order to apply the code to the practical designing, authors settled a quasi-component two-dimensional axisymmetric model and a loading cycle (500 cycles). Then, an inelastic analysis and its integrity evaluation are carried out

  18. Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

    International Nuclear Information System (INIS)

    Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.

    1987-01-01

    ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

  19. Evaluation of large girth LDPC codes for PMD compensation by turbo equalization.

    Science.gov (United States)

    Minkov, Lyubomir L; Djordjevic, Ivan B; Xu, Lei; Wang, Ting; Kueppers, Franko

    2008-08-18

    Large-girth quasi-cyclic LDPC codes have been experimentally evaluated for use in PMD compensation by turbo equalization for a 10 Gb/s NRZ optical transmission system, and observing one sample per bit. Net effective coding gain improvement for girth-10, rate 0.906 code of length 11936 over maximum a posteriori probability (MAP) detector for differential group delay of 125 ps is 6.25 dB at BER of 10(-6). Girth-10 LDPC code of rate 0.8 outperforms the girth-10 code of rate 0.906 by 2.75 dB, and provides the net effective coding gain improvement of 9 dB at the same BER. It is experimentally determined that girth-10 LDPC codes of length around 15000 approach channel capacity limit within 1.25 dB.

  20. Fetal cerebro-placental ratio and adverse perinatal outcome: systematic review and meta-analysis of the association and diagnostic performance.

    Science.gov (United States)

    Nassr, Ahmed Abobakr; Abdelmagied, Ahmed M; Shazly, Sherif A M

    2016-03-01

    The objective of this meta-analysis is to assess the value of fetal cerebro-placental Doppler ratio (CPR) in predicting adverse perinatal outcome in pregnancies with fetal growth restriction (FGR). Three databases were used: MEDLINE, EMBASE (with online Ovid interface) and SCOPUS and studies from inception to April 2015 were included. Studies that reported perinatal outcomes of fetuses at risk of FGR or sonographically diagnosed FGR that were evaluated with CPR were considered eligible. Perinatal outcomes include cesarean section (CS) for fetal distress, APGAR scores at 5 min, neonatal complications and admission to neonatal intensive care unit (NICU). Pooled data were expressed as odds ratio (OR) and confidence intervals (CI), and the summary receiver operating characteristic (SROC) curve was used to illustrate the diagnostic accuracy of CPR. Seven studies were eligible (1428 fetuses). Fetuses with abnormal CPR were at higher risk of CS for fetal distress (OR=4.49, 95% CI [1.63, 12.42]), lower APGAR scores (OR=4.01, 95% CI [2.65, 6.08]), admission to NICU (OR=9.65, 95% CI [3.02, 30.85]), and neonatal complications (OR=11.00, 95% [3.64, 15.37]) than fetuses who had normal CPR. These risks were higher among studies that included fetuses diagnosed with FGR than fetuses at risk of FGR. Abnormal CPR had higher diagnostic accuracy for adverse perinatal outcomes among "sonographically diagnosed FGR" studies than "at risk of FGR" studies. Abnormal CPR is associated with substantial risk of adverse perinatal outcomes. The test seems to be particularly useful for follow up of fetuses with sonographically diagnosed FGR.

  1. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  2. Coded Statutory Data Sets for Evaluation of Public Health Law

    Science.gov (United States)

    Costich, Julia Field

    2012-01-01

    Background and objectives: The evaluation of public health law requires reliable accounts of underlying statutes and regulations. States often enact public health-related statutes with nonuniform provisions, and variation in the structure of state legal codes can foster inaccuracy in evaluating the impact of specific categories of law. The optimal…

  3. A comparative evaluation of NDR and PSAR using the CASMO-3/MASTER code system

    International Nuclear Information System (INIS)

    Sim, Jeoung Hun; Kim, Han Gon

    2009-01-01

    In order to validate nuclear design data such as the nuclear design report (NDR) and data in preliminary (or final) safety analysis report (PSAR/FSAR) and to use data for the conceptual design of new plants, the CASMO-3/MASTER code system is selected as utility code. The nuclear design of OPR1000 and APR1400 is performed with the DIT/ROCS code system. In contrast with this design code system, the accuracy of CASMO- 3/MASTER code system has not been verified. Relatively little design data has been calculated by the CASMO-3/MASTER code system for OPR1000 and APR1400 and a bias system has not been developed yet. As such, validation of the performance of the CASMO- 3/MASTER code system is necessary. In order to validate the performance of the CASMO- 3/MASTER code system and to develop a calculation methodology, a comparative evaluation with NDR of Ulchin unit 4, cycle 1(U4C1) and the PSAR of Shinkori units 3 and 4 is carried out. The results of this evaluation are presented in this paper

  4. Computer codes for evaluation of control room habitability (HABIT)

    International Nuclear Information System (INIS)

    Stage, S.A.

    1996-06-01

    This report describes the Computer Codes for Evaluation of Control Room Habitability (HABIT). HABIT is a package of computer codes designed to be used for the evaluation of control room habitability in the event of an accidental release of toxic chemicals or radioactive materials. Given information about the design of a nuclear power plant, a scenario for the release of toxic chemicals or radionuclides, and information about the air flows and protection systems of the control room, HABIT can be used to estimate the chemical exposure or radiological dose to control room personnel. HABIT is an integrated package of several programs that previously needed to be run separately and required considerable user intervention. This report discusses the theoretical basis and physical assumptions made by each of the modules in HABIT and gives detailed information about the data entry windows. Sample runs are given for each of the modules. A brief section of programming notes is included. A set of computer disks will accompany this report if the report is ordered from the Energy Science and Technology Software Center. The disks contain the files needed to run HABIT on a personal computer running DOS. Source codes for the various HABIT routines are on the disks. Also included are input and output files for three demonstration runs

  5. Licensing in BE system code calculations. Applications and uncertainty evaluation by CIAU method

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco

    2007-01-01

    The evaluation of uncertainty constitutes the necessary supplement of Best Estimate (BE) calculations performed to understand accident scenarios in water cooled nuclear reactors. The needs come from the imperfection of computational tools on the one side and from the interest in using such tool to get more precise evaluation of safety margins. In the present paper the approaches to uncertainty are outlined and the CIAU (Code with capability of Internal Assessment of Uncertainty) method proposed by the University of Pisa is described including ideas at the basis and results from applications. Two approaches are distinguished that are characterized as 'propagation of code input uncertainty' and 'propagation of code output errors'. For both methods, the thermal-hydraulic code is at the centre of the process of uncertainty evaluation: in the former case the code itself is adopted to compute the error bands and to propagate the input errors, in the latter case the errors in code application to relevant measurements are used to derive the error bands. The CIAU method exploits the idea of the 'status approach' for identifying the thermal-hydraulic conditions of an accident in any Nuclear Power Plant (NPP). Errors in predicting such status are derived from the comparison between predicted and measured quantities and, in the stage of the application of the method, are used to compute the uncertainty. (author)

  6. Evaluation of the computer code system RADHEAT-V4 by analysing benchmark problems on radiation shielding

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Naito, Yoshitaka

    1990-11-01

    A computer code system RADHEAT-V4 has been developed for safety evaluation on radiation shielding of nuclear fuel facilities. To evaluate the performance of the code system, 18 benchmark problem were selected and analysed. Evaluated radiations are neutron and gamma-ray. Benchmark problems consist of penetration, streaming and skyshine. The computed results show more accurate than those by the Sn codes ANISN and DOT3.5 or the Monte Carlo code MORSE. Big core memory and many times I/O are, however, required for RADHEAT-V4. (author)

  7. Evaluation of the RELAP4/MOD6 thermal-hydraulic code

    International Nuclear Information System (INIS)

    Haigh, W.S.; Margolis, S.G.; Rice, R.E.

    1978-01-01

    The NRC RELAP4/MOD6 computer code was recently released to the public for use in thermal-hydraulic analysis. This code has a unique new capability permitting analysis of both the blowdown and reflood portions of a postulated pressurized water reactor (PWR) loss-of-coolant accident (LOCA). A principal code evaluation objective is to assess the accuracy of the code for computing LOCA behavior over a wide range of system sizes and scaling concepts. The scales of interest include all LOCA experiments and will ultimately encompass full-sized PWR systems for which no experiments or data are available. Quantitative assessment of the accuracy of the code when it is applied to large PWR systems is still in the future. With RELAP4/MOD6, however, a technique has been demonstrated for using results derived from small-scale blowdown and reflood experiments to predict the accuracy of calculations for similar experiments of significantly different scale or component size. This demonstration is considered a first step in establishing confidence levels for the accuracy of calculations of a postulated LOCA

  8. Performance and Complexity Evaluation of Iterative Receiver for Coded MIMO-OFDM Systems

    Directory of Open Access Journals (Sweden)

    Rida El Chall

    2016-01-01

    Full Text Available Multiple-input multiple-output (MIMO technology in combination with channel coding technique is a promising solution for reliable high data rate transmission in future wireless communication systems. However, these technologies pose significant challenges for the design of an iterative receiver. In this paper, an efficient receiver combining soft-input soft-output (SISO detection based on low-complexity K-Best (LC-K-Best decoder with various forward error correction codes, namely, LTE turbo decoder and LDPC decoder, is investigated. We first investigate the convergence behaviors of the iterative MIMO receivers to determine the required inner and outer iterations. Consequently, the performance of LC-K-Best based receiver is evaluated in various LTE channel environments and compared with other MIMO detection schemes. Moreover, the computational complexity of the iterative receiver with different channel coding techniques is evaluated and compared with different modulation orders and coding rates. Simulation results show that LC-K-Best based receiver achieves satisfactory performance-complexity trade-offs.

  9. Analysis and application of ratcheting evaluation procedure of Japanese high temperature design code DDS

    International Nuclear Information System (INIS)

    Lee, H. Y.; Kim, J. B.; Lee, J. H.

    2002-01-01

    In this study, the evaluation procedure of Japanese DDS code which was recently developed to assess the progressive inelastic deformation occurring under repetition of secondary stresses was analyzed and the evaluation results according to DDS was compared those of the thermal ratchet structural test carried out by KAERI to analyze the conservativeness of the code. The existing high temperature codes of US ASME-NH and French RCC-MR suggest the limited ratcheting procedures for only the load cases of cyclic secondary stresses under primary stresses. So they are improper to apply to the actual ratcheting problem which can occur under cyclic secondary membrane stresses due to the movement of hot free surface for the pool type LMR. DDS provides explicitly an analysis procedure of ratcheting due to moving thermal gradients near hot free surface. A comparison study was carried out between the results by the design code of DDS and by the structural test to investigate the conservativeness of DDS code, which showed that the evaluation results by DDS were in good agreement with those of the structural test

  10. Evaluation of an electrocardiogram on QR code.

    Science.gov (United States)

    Nakayama, Masaharu; Shimokawa, Hiroaki

    2013-01-01

    An electrocardiogram (ECG) is an indispensable tool to diagnose cardiac diseases, such as ischemic heart disease, myocarditis, arrhythmia, and cardiomyopathy. Since ECG patterns vary depend on patient status, it is also used to monitor patients during treatment and comparison with ECGs with previous results is important for accurate diagnosis. However, the comparison requires connection to ECG data server in a hospital and the availability of data connection among hospitals is limited. To improve the portability and availability of ECG data regardless of server connection, we here introduce conversion of ECG data into 2D barcodes as text data and decode of the QR code for drawing ECG with Google Chart API. Fourteen cardiologists and six general physicians evaluated the system using iPhone and iPad. Overall, they were satisfied with the system in usability and accuracy of decoded ECG compared to the original ECG. This new coding system may be useful in utilizing ECG data irrespective of server connections.

  11. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-01-01

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

  12. Evaluation of crack-like flaw in Japanese fitness-for-service code for nuclear power plant components

    International Nuclear Information System (INIS)

    Kashima, Koichi

    2003-01-01

    For evaluation of faults detected at nuclear appliances, establishment of fitness-for-service code in Japan is focused by most of peoples. The code is a management rule to keep features of the appliances under supplying operation to their constant safe level and is a rule composing a pair with design rule. The codes for nuclear power generation facilities-rules of fitness-for-service for nuclear power plants were issued on May, 2002, by the Japan Society of Mechanical Engineering (JSME), which was added on October, 2002, by its inspection code, for its amendment. Under such states, Japan Government is proceeding on establishment of the fitness-for-service code in Japan on a base of the private rule. Here were introduced present state and tasks on content of crack-like flaw evaluation on the code under an example of the private rule of JSME, which is composed of three items of inspection, evaluation, and recovery and exchange. The evaluation of defects consists of 1) the first step of evaluation of defects and 2) the second step of evaluation of defects. The first step determines the size of defect by modeling form. When the size of defect is smaller than the evaluation criterion, the appliances can be used unconditionally. However, its size is larger than the evaluation criterion, the appliances have to be evaluated by the second step. When the estimated defects size at end of evaluation period is smaller than the permissible value, the appliances can be used within the evaluation period. But, if its size is larger than the permissible value, the appliances have to be recovered and exchanged. Modeling, evaluation criterion, evaluation of destruction, safety standards and future problems are described. (S.Y.)

  13. Computer code for quantitative ALARA evaluations

    International Nuclear Information System (INIS)

    Voilleque, P.G.

    1984-01-01

    A FORTRAN computer code has been developed to simplify the determination of whether dose reduction actions meet the as low as is reasonably achievable (ALARA) criterion. The calculations are based on the methodology developed for the Atomic Industrial Forum. The code is used for analyses of eight types of dose reduction actions, characterized as follows: reduce dose rate, reduce job frequency, reduce productive working time, reduce crew size, increase administrative dose limit for the task, and increase the workers' time utilization and dose utilization through (a) improved working conditions, (b) basic skill training, or (c) refresher training for special skills. For each type of action, two analysis modes are available. The first is a generic analysis in which the program computes potential benefits (in dollars) for a range of possible improvements, e.g., for a range of lower dose rates. Generic analyses are most useful in the planning stage and for evaluating the general feasibility of alternative approaches. The second is a specific analysis in which the potential annual benefits of a specific level of improvement and the annual implementation cost are compared. The potential benefits reflect savings in operational and societal costs that can be realized if occupational radiation doses are reduced. Because the potential benefits depend upon many variables which characterize the job, the workplace, and the workers, there is no unique relationship between the potential dollar savings and the dose savings. The computer code permits rapid quantitative analyses of alternatives and is a tool that supplements the health physicist's professional judgment. The program output provides a rational basis for decision-making and a record of the assumptions employed

  14. Evaluation of ATLAS 100% DVI Line Break Using TRACE Code

    International Nuclear Information System (INIS)

    Huh, Byung Gil; Bang, Young Seok; Cheong, Ae Ju; Woo, Sweng Woong

    2011-01-01

    ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is an integral effect test facility in KAERI. It had installed completely to simulate the accident for the OPR1000 and the APR1400 in 2005. After then, several tests for LBLOCA, DVI line break have been performed successfully to resolve the safety issues of the APR1400. Especially, a DVI line break is considered as another spectrum among the SBLOCAs in APR1400 because the DVI line is directly connected to the reactor vessel and the thermal hydraulic behaviors are expected to be different from those for the cold leg injection. However, there are not enough experimental data for the DVI line break. Therefore, integral effect data for the DVI line break of ATLAS is very useful and available for an improvement and validation of safety codes. For the DVI line break in ATLAS, several analyses using MARS and RELAP codes were performed in the ATLAS DSP (Domestic Standard Problem) meetings. However, TRACE code has still not used to simulate a DVI line break. TRACE code has developed as the unified code for the reactor thermal hydraulic analyses in USNRC. In this study, the 100% DVI line break in ATLAS was evaluated by TRACE code. The objectives of this study are to identify the prediction capability of TRACE code for the major thermal hydraulic phenomena of a DVI line break in ATLAS

  15. DOZIM - evaluation dose code for nuclear accident

    International Nuclear Information System (INIS)

    Oprea, I.; Musat, D.; Ionita, I.

    2008-01-01

    During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 10 2 to 2 x 10 4 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on

  16. A model of R-D performance evaluation for Rate-Distortion-Complexity evaluation of H.264 video coding

    DEFF Research Database (Denmark)

    Wu, Mo; Forchhammer, Søren

    2007-01-01

    This paper considers a method for evaluation of Rate-Distortion-Complexity (R-D-C) performance of video coding. A statistical model of the transformed coefficients is used to estimate the Rate-Distortion (R-D) performance. A model frame work for rate, distortion and slope of the R-D curve for inter...... and intra frame is presented. Assumptions are given for analyzing an R-D model for fast R-D-C evaluation. The theoretical expressions are combined with H.264 video coding, and confirmed by experimental results. The complexity frame work is applied to the integer motion estimation....

  17. GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications

    International Nuclear Information System (INIS)

    Glaeser, H.

    2008-01-01

    During the recent years, an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from best estimate thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour

  18. The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

    Science.gov (United States)

    Iwamoto, O.; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

    2016-01-01

    A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

  19. The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, O., E-mail: iwamoto.osamu@jaea.go.jp; Iwamoto, N.; Kunieda, S.; Minato, F.; Shibata, K.

    2016-01-15

    A computer code system, CCONE, was developed for nuclear data evaluation within the JENDL project. The CCONE code system integrates various nuclear reaction models needed to describe nucleon, light charged nuclei up to alpha-particle and photon induced reactions. The code is written in the C++ programming language using an object-oriented technology. At first, it was applied to neutron-induced reaction data on actinides, which were compiled into JENDL Actinide File 2008 and JENDL-4.0. It has been extensively used in various nuclear data evaluations for both actinide and non-actinide nuclei. The CCONE code has been upgraded to nuclear data evaluation at higher incident energies for neutron-, proton-, and photon-induced reactions. It was also used for estimating β-delayed neutron emission. This paper describes the CCONE code system indicating the concept and design of coding and inputs. Details of the formulation for modelings of the direct, pre-equilibrium and compound reactions are presented. Applications to the nuclear data evaluations such as neutron-induced reactions on actinides and medium-heavy nuclei, high-energy nucleon-induced reactions, photonuclear reaction and β-delayed neutron emission are mentioned.

  20. Evaluation of Yonggwang unit 4 cycle 5 using SPNOVA code

    International Nuclear Information System (INIS)

    Choi, Y. S.; Cha, K. H.; Lee, E. K.; Park, M. K.

    2004-01-01

    Core follow calculation of Yonggwang (YGN) unit 4 cycle 5 is performed to evaluate SPNOVA code if it can be applicable or not to Korean standard nuclear power plant (KSNP). SPNOVA code consists of BEPREPN and ANC code to represent incore detector and neutronics model, respectively. SPNOVA core deflection model is compared and verified with ANC depletion results in terms of critical boron concentration (CBC), peaking factor (Fq) and radial power distribution. In YGN4, SPNOVA predicts 30 ppm lower than that of ROCS predicting CBC. Fq and radial power distribution behavior of SPNOVA calculation have conservatively higher than those of ROCS predicting values. And also SPNOVA predicting results are compared with measurement data from snapshot and CECOR core calculation. It is reasonable to accept SPNOVA to analyze KSNP. The model of SPNOVA for KSNP will be used to develop the brand-new incore detector of platinum and vanadium

  1. Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test

    International Nuclear Information System (INIS)

    Auria, F.D.; Oriolo, F.; Leonardi, M.; Paci, S.

    1995-01-01

    Quantitative evaluation of code uncertainties is a necessary step in the code assessment process, above all if best-estimate codes are utilised for licensing purposes. Aiming at quantifying the code accuracy, an integral methodology based on the Fast Fourier Transform (FFT) has been developed at the University of Pisa (DCMN) and has been already applied to several calculations related to primary system test analyses. This paper deals with the first application of the FFT based methodology to containment code calculations based on a hydrogen mixing and distribution test performed in the NUPEC (Nuclear Power Engineering Corporation) facility. It is referred to pre-test and post-test calculations submitted for the International Standard Problem (ISP) n. 35. This is a blind exercise, simulating the effects of steam injection and spray behaviour on gas distribution and mixing. The result of the application of this methodology to nineteen selected variables calculated by ten participants are here summarized, and the comparison (where possible) of the accuracy evaluated for the pre-test and for the post-test calculations of a same user is also presented. (author)

  2. Elastic creep-fatigue evaluation for ASME [American Society of Mechanical Engineers] code

    International Nuclear Information System (INIS)

    Severud, L.K.; Winkel, B.V.

    1987-02-01

    Reassessment of past ASME N-47 creep-fatigue rules have been under way by committee members. The new proposed elastic creep-fatigue methods are easier to apply than those previously in the code case. They also provide a wider range of practical application while still providing conservative assessments. It is expected that new N-47 code rules for elastic creep-fatigue evaluation will be adopted in the near future

  3. Fresh Prime Codes Evaluation for Synchronous PPM and OPPM Signaling for Optical CDMA Networks

    Science.gov (United States)

    Karbassian, M. Massoud; Ghafouri-Shiraz, H.

    2007-06-01

    In this paper, we have proposed a novel prime spreading sequence family hereby referred to as “Double-Padded Modified Prime Code (DPMPC)” for direct-detection synchronous optical code-division multiple-access (OCDMA) networks. The new code is applied to both pulse-position and overlapping pulse-position modulation CDMA networks, and their performances were evaluated and compared with existing prime codes family. In addition, we have analyzed the system throughput and also introduced a new interference cancellation technique which significantly improves the bit error probability of OCDMA networks.

  4. Integrity evaluation for stud female threads on pressure vessel according to ASME code using FEM

    International Nuclear Information System (INIS)

    Kim, Moon Young; Chung, Nam Yong

    2003-01-01

    The extension of design life among power plants is increasingly becoming a world-wide trend. Kori no.1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts for man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety for helical-coil method which is used according to code case-N-496-1. From analysis results, we found that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10% by manufacture integrity criteria and 2/3 yield strength criteria on ASME code. It was also confirmed that the helical-coil repair method would be safe

  5. Game-Coding Workshops in New Zealand Public Libraries: Evaluation of a Pilot Project

    Science.gov (United States)

    Bolstad, Rachel

    2016-01-01

    This report evaluates a game coding workshop offered to young people and adults in seven public libraries round New Zealand. Participants were taken step by step through the process of creating their own simple 2D videogame, learning the basics of coding, computational thinking, and digital game design. The workshops were free and drew 426 people…

  6. SGV: a code to evaluate plasma reaction rates to a specified accuracy

    Energy Technology Data Exchange (ETDEWEB)

    Devoto, R.S.; Hanson, J.D.

    1978-09-22

    A FORTRAN code to evaluate binary reaction rates (sigmav) for a plasma to a specified accuracy is described. Distribution functions permitted are (1) two Maxwellian species at different temperatures, (2) beam-Maxwellian, (3) cold gas with Maxwellian, and (4) beam-plasma with mirror distribution of the form f(v) varies as f(v) M (cos theta). Several functional forms are permitted for f(v) and M(cos theta). Cross-section subroutines for a number of interactions involving hydrogen, helium, and electrons are included, as is a routine allowing input of numerical data. The code is written as a subroutine to allow ready incorporation into larger plasma codes.

  7. An In vitro evaluation of the reliability of QR code denture labeling technique.

    Science.gov (United States)

    Poovannan, Sindhu; Jain, Ashish R; Krishnan, Cakku Jalliah Venkata; Chandran, Chitraa R

    2016-01-01

    Positive identification of the dead after accidents and disasters through labeled dentures plays a key role in forensic scenario. A number of denture labeling methods are available, and studies evaluating their reliability under drastic conditions are vital. This study was conducted to evaluate the reliability of QR (Quick Response) Code labeled at various depths in heat-cured acrylic blocks after acid treatment, heat treatment (burns), and fracture in forensics. It was an in vitro study. This study included 160 specimens of heat-cured acrylic blocks (1.8 cm × 1.8 cm) and these were divided into 4 groups (40 samples per group). QR Codes were incorporated in the samples using clear acrylic sheet and they were assessed for reliability under various depths, acid, heat, and fracture. Data were analyzed using Chi-square test, test of proportion. The QR Code inclusion technique was reliable under various depths of acrylic sheet, acid (sulfuric acid 99%, hydrochloric acid 40%) and heat (up to 370°C). Results were variable with fracture of QR Code labeled acrylic blocks. Within the limitations of the study, by analyzing the results, it was clearly indicated that the QR Code technique was reliable under various depths of acrylic sheet, acid, and heat (370°C). Effectiveness varied in fracture and depended on the level of distortion. This study thus suggests that QR Code is an effective and simpler denture labeling method.

  8. Towards high dynamic range extensions of HEVC: subjective evaluation of potential coding technologies

    Science.gov (United States)

    Hanhart, Philippe; Řeřábek, Martin; Ebrahimi, Touradj

    2015-09-01

    This paper reports the details and results of the subjective evaluations conducted at EPFL to evaluate the responses to the Call for Evidence (CfE) for High Dynamic Range (HDR) and Wide Color Gamut (WCG) Video Coding issued by Moving Picture Experts Group (MPEG). The CfE on HDR/WCG Video Coding aims to explore whether the coding efficiency and/or the functionality of the current version of HEVC standard can be signi_cantly improved for HDR and WCG content. In total, nine submissions, five for Category 1 and four for Category 3a, were compared to the HEVC Main 10 Profile based Anchor. More particularly, five HDR video contents, compressed at four bit rates by each proponent responding to the CfE, were used in the subjective evaluations. Further, the side-by-side presentation methodology was used for the subjective experiment to discriminate small differences between the Anchor and proponents. Subjective results shows that the proposals provide evidence that the coding efficiency can be improved in a statistically noticeable way over MPEG CfE Anchors in terms of perceived quality within the investigated content. The paper further benchmarks the selected objective metrics based on their correlations with the subjective ratings. It is shown that PSNR-DE1000, HDRVDP- 2, and PSNR-Lx can reliably detect visible differences between the proposed encoding solutions and current HEVC standard.

  9. A clinical evaluation of placental growth factor in routine practice in high-risk women presenting with suspected pre-eclampsia and/or fetal growth restriction.

    Science.gov (United States)

    Ormesher, L; Johnstone, E D; Shawkat, E; Dempsey, A; Chmiel, C; Ingram, E; Higgins, L E; Myers, J E

    2018-03-13

    To evaluate the use of plasma Placental Growth Factor (PlGF), recommended by the recent NICE guidance, in women with suspected pre-eclampsia (PE) and/or fetal growth restriction (FGR). Non-randomised prospective clinical evaluation study in high-risk antenatal clinics in a tertiary maternity unit. PlGF testing was performed in addition to routine clinical assessment in 260 women >20 weeks' gestation with chronic disease (hypertension, renal disease ± diabetes) with a change in maternal condition or in women with suspected FGR to determine the impact on clinical management. Results were revealed and standardised care pathways followed. Outcome of pregnancies with a low PlGF (women had an adverse outcome (PE/birthweight women with PlGF 14 days. The PlGF result altered clinical management (surveillance or timing of birth) in 196/260 (75.4%) cases. Alternative PlGF thresholds did not significantly improve diagnostic performance. Our evaluation confirms the value of PlGF as a diagnostic tool for placental dysfunction. However, low PlGF in isolation should not trigger iatrogenic delivery. Further research linking placental pathology, maternal disease and maternal PlGF levels is urgently needed before this test can be implemented in routine clinical practice. Copyright © 2018. Published by Elsevier B.V.

  10. Trends in EFL Technology and Educational Coding: A Case Study of an Evaluation Application Developed on LiveCode

    Science.gov (United States)

    Uehara, Suwako; Noriega, Edgar Josafat Martinez

    2016-01-01

    The availability of user-friendly coding software is increasing, yet teachers might hesitate to use this technology to develop for educational needs. This paper discusses studies related to technology for educational uses and introduces an evaluation application being developed. Through questionnaires by student users and open-ended discussion by…

  11. LDGM Codes for Channel Coding and Joint Source-Channel Coding of Correlated Sources

    Directory of Open Access Journals (Sweden)

    Javier Garcia-Frias

    2005-05-01

    Full Text Available We propose a coding scheme based on the use of systematic linear codes with low-density generator matrix (LDGM codes for channel coding and joint source-channel coding of multiterminal correlated binary sources. In both cases, the structures of the LDGM encoder and decoder are shown, and a concatenated scheme aimed at reducing the error floor is proposed. Several decoding possibilities are investigated, compared, and evaluated. For different types of noisy channels and correlation models, the resulting performance is very close to the theoretical limits.

  12. ENDF-UTILITY-CODES, codes to check and standardize data in the Evaluated Nuclear Data File (ENDF)

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: The ENDF Utility Codes include 9 codes to check and standardize data in the Evaluated Nuclear Data File (ENDF). Four programs of this release, GETMAT, LISTEF, PLOTEF and SETMDC are no more maintained since release 6.13. The suite of ENDF utility codes includes: - CHECKR (version 7.01) is a program for checking that an evaluated data file conforms to the ENDF format. - FIZCON (version 7.02) is a program for checking that an evaluated data file has valid data and conforms to recommended procedures. - GETMAT (version 6.13) is designed to retrieve one or more materials from an ENDF formatted data file. The output will contain only the selected materials. - INTER (version 7.01) calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. - LISTEF (version 6.13) is designed to produce summary and annotated listings of a data file in either ENDF-6 or ENDF-5 format. - PLOTEF (version 6.13) is designed to produce graphical displays of a data file in either ENDF-5 or ENDF-6 format. The form of graphical output depends on the graphical devices available at the installation where this code will be used. - PSYCHE (version 7.02) is a program for checking the physics content of an evaluated data file. It can recognise the difference between ENDF-5 or ENDF-6 formats and performs its tests accordingly. - SETMDC (version 6.13) is a utility program that converts the source decks of programs to different computers (DOS, UNIX, LINUX, VMS, Windows). - STANEF (version 7.01) performs bookkeeping operations on a data file containing one or more material evaluations in ENDF format. The version 7.02 of the ENDF Utility Codes corrects all bugs reported to NNDC as of April 1, 2005 and supersedes all previous releases. Three codes CHECKR, STANEF, and INTER were actually ported from the 7.01 release

  13. Methods of evaluating the effects of coding on SAR data

    Science.gov (United States)

    Dutkiewicz, Melanie; Cumming, Ian

    1993-01-01

    It is recognized that mean square error (MSE) is not a sufficient criterion for determining the acceptability of an image reconstructed from data that has been compressed and decompressed using an encoding algorithm. In the case of Synthetic Aperture Radar (SAR) data, it is also deemed to be insufficient to display the reconstructed image (and perhaps error image) alongside the original and make a (subjective) judgment as to the quality of the reconstructed data. In this paper we suggest a number of additional evaluation criteria which we feel should be included as evaluation metrics in SAR data encoding experiments. These criteria have been specifically chosen to provide a means of ensuring that the important information in the SAR data is preserved. The paper also presents the results of an investigation into the effects of coding on SAR data fidelity when the coding is applied in (1) the signal data domain, and (2) the image domain. An analysis of the results highlights the shortcomings of the MSE criterion, and shows which of the suggested additional criterion have been found to be most important.

  14. Evaluation Codes from Order Domain Theory

    DEFF Research Database (Denmark)

    Andersen, Henning Ejnar; Geil, Hans Olav

    2008-01-01

    bound is easily extended to deal with any generalized Hamming weights. We interpret our methods into the setting of order domain theory. In this way we fill in an obvious gap in the theory of order domains. [28] T. Shibuya and K. Sakaniwa, A Dual of Well-Behaving Type Designed Minimum Distance, IEICE......The celebrated Feng-Rao bound estimates the minimum distance of codes defined by means of their parity check matrices. From the Feng-Rao bound it is clear how to improve a large family of codes by leaving out certain rows in their parity check matrices. In this paper we derive a simple lower bound...... on the minimum distance of codes defined by means of their generator matrices. From our bound it is clear how to improve a large family of codes by adding certain rows to their generator matrices. The new bound is very much related to the Feng-Rao bound as well as to Shibuya and Sakaniwa's bound in [28]. Our...

  15. An evaluation of the effect of JPEG, JPEG2000, and H.264/AVC on CQR codes decoding process

    Science.gov (United States)

    Vizcarra Melgar, Max E.; Farias, Mylène C. Q.; Zaghetto, Alexandre

    2015-02-01

    This paper presents a binarymatrix code based on QR Code (Quick Response Code), denoted as CQR Code (Colored Quick Response Code), and evaluates the effect of JPEG, JPEG2000 and H.264/AVC compression on the decoding process. The proposed CQR Code has three additional colors (red, green and blue), what enables twice as much storage capacity when compared to the traditional black and white QR Code. Using the Reed-Solomon error-correcting code, the CQR Code model has a theoretical correction capability of 38.41%. The goal of this paper is to evaluate the effect that degradations inserted by common image compression algorithms have on the decoding process. Results show that a successful decoding process can be achieved for compression rates up to 0.3877 bits/pixel, 0.1093 bits/pixel and 0.3808 bits/pixel for JPEG, JPEG2000 and H.264/AVC formats, respectively. The algorithm that presents the best performance is the H.264/AVC, followed by the JPEG2000, and JPEG.

  16. Evaluating QR Code Case Studies Using a Mobile Learning Framework

    Science.gov (United States)

    Rikala, Jenni

    2014-01-01

    The aim of this study was to evaluate the feasibility of Quick Response (QR) codes and mobile devices in the context of Finnish basic education. The feasibility was analyzed through a mobile learning framework, which includes the core characteristics of mobile learning. The study is part of a larger research where the aim is to develop a…

  17. Material Performance of Fully-Ceramic Micro-Encapsulated Fuel under Selected LWR Design Basis Scenarios: Final Report

    International Nuclear Information System (INIS)

    Boer, B.; Sen, R.S.; Pope, M.A.; Ougouag, A.M.

    2011-01-01

    The extension to LWRs of the use of Deep-Burn coated particle fuel envisaged for HTRs has been investigated. TRISO coated fuel particles are used in Fully-Ceramic Microencapsulated (FCM) fuel within a SiC matrix rather than the graphite of HTRs. TRISO particles are well characterized for uranium-fueled HTRs. However, operating conditions of LWRs are different from those of HTRs (temperature, neutron energy spectrum, fast fluence levels, power density). Furthermore, the time scales of transient core behavior during accidents are usually much shorter and thus more severe in LWRs. The PASTA code was updated for analysis of stresses in coated particle FCM fuel. The code extensions enable the automatic use of neutronic data (burnup, fast fluence as a function of irradiation time) obtained using the DRAGON neutronics code. An input option for automatic evaluation of temperature rise during anticipated transients was also added. A new thermal model for FCM was incorporated into the code; so-were updated correlations (for pyrocarbon coating layers) suitable to estimating dimensional changes at the high fluence levels attained in LWR DB fuel. Analyses of the FCM fuel using the updated PASTA code under nominal and accident conditions show: (1) Stress levels in SiC-coatings are low for low fission gas release (FGR) fractions of several percent, as based on data of fission gas diffusion in UO 2 kernels. However, the high burnup level of LWR-DB fuel implies that the FGR fraction is more likely to be in the range of 50-100%, similar to Inert Matrix Fuels (IMFs). For this range the predicted stresses and failure fractions of the SiC coating are high for the reference particle design (500 (micro)mm kernel diameter, 100 (micro)mm buffer, 35 (micro)mm IPyC, 35 (micro)mm SiC, 40 (micro)mm OPyC). A conservative case, assuming 100% FGR, 900K fuel temperature and 705 MWd/kg (77% FIMA) fuel burnup, results in a 8.0 x 10 -2 failure probability. For a 'best-estimate' FGR fraction of 50

  18. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  19. Evaluation of SPACE code for simulation of inadvertent opening of spray valve in Shin Kori unit 1

    International Nuclear Information System (INIS)

    Kim, Seyun; Youn, Bumsoo

    2013-01-01

    SPACE code is expected to be applied to the safety analysis for LOCA (Loss of Coolant Accident) and Non-LOCA scenarios. SPACE code solves two-fluid, three-field governing equations and programmed with C++ computer language using object-oriented concepts. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code

  20. Modern Nuclear Data Evaluation with the TALYS Code System

    Science.gov (United States)

    Koning, A. J.; Rochman, D.

    2012-12-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: "Total" Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  1. Modern Nuclear Data Evaluation with the TALYS Code System

    International Nuclear Information System (INIS)

    Koning, A.J.; Rochman, D.

    2012-01-01

    This paper presents a general overview of nuclear data evaluation and its applications as developed at NRG, Petten. Based on concepts such as robustness, reproducibility and automation, modern calculation tools are exploited to produce original nuclear data libraries that meet the current demands on quality and completeness. This requires a system which comprises differential measurements, theory development, nuclear model codes, resonance analysis, evaluation, ENDF formatting, data processing and integral validation in one integrated approach. Software, built around the TALYS code, will be presented in which all these essential nuclear data components are seamlessly integrated. Besides the quality of the basic data and its extensive format testing, a second goal lies in the diversity of processing for different type of users. The implications of this scheme are unprecedented. The most important are: 1. Complete ENDF-6 nuclear data files, in the form of the TENDL library, including covariance matrices, for many isotopes, particles, energies, reaction channels and derived quantities. All isotopic data files are mutually consistent and are supposed to rival those of the major world libraries. 2. More exact uncertainty propagation from basic nuclear physics to applied (reactor) calculations based on a Monte Carlo approach: “Total” Monte Carlo (TMC), using random nuclear data libraries. 3. Automatic optimization in the form of systematic feedback from integral measurements back to the basic data. This method of work also opens a new way of approaching the analysis of nuclear applications, with consequences in both applied nuclear physics and safety of nuclear installations, and several examples are given here. This applied experience and feedback is integrated in a final step to improve the quality of the nuclear data, to change the users vision and finally to orchestrate their integration into simulation codes.

  2. Forest gene conservation from the perspective of the international community

    Science.gov (United States)

    M. Hosny El-Lakany

    2017-01-01

    conservation of forest genetic resources (FGR). After presenting internationally adopted definitions of some terms related to FGR, the characteristics of the current state of FGR conservation from a global perspective are summarized. Many international and regional organizations and institutions are engaged in the conservation of FGR at degrees ranging from...

  3. PP043. Oxidative stress in the maternal body also affects the fetus in preeclamptic women with fetal growth restriction.

    Science.gov (United States)

    Watanabe, Kazushi; Iwasaki, Ai; Mori, Toshitaka; Kimura, Chiharu; Matsushita, Hiroshi; Shinohara, Koichi; Wakatsuki, Akihiko

    2013-04-01

    The purpose of the present study was to determine whether oxidative stress occurring in the maternal body also affects the fetus in preeclamptic women with FGR. We ∥@consecutively recruited 17 preeclamptic women with FGR, 16 preeclamptic women without FGR, and 16 healthy pregnant women with uncomplicated pregnancy. We measured concentrations of derivatives of reactive oxygen metabolites (d-ROMs) as a marker of oxygen free radicals in a maternal vein, umbilical artery, and umbilical vein. ∥@Maternal d-ROM levels were higher in preeclamptic groups compared to the control group. Umbilical artery and vein d-ROM levels were elevated in preeclamptic women with FGR compared to the control group. Umbilical artery d-ROM levels were significantly higher than in the vein in preeclamptic women with FGR, but not in those without FGR. Umbilical arterial blood pH was significantly lower in preeclamptic women with FGR. The partial pressure of oxygen (PaO2) in umbilical arterial blood tended to be lower in preeclamptic women with FGR (p=0.08). The partial pressure of carbon dioxide (PaCO2) in umbilical arterial blood was significantly higher in preeclamptic women with FGR. These results indicate that oxidative stress occurring in the maternal body also affects the fetus in preeclamptic women with FGR. Copyright © 2013. Published by Elsevier B.V.

  4. Probabilistic evaluation of design S-N curve and reliability assessment of ASME code-based evaluation

    International Nuclear Information System (INIS)

    Zhao Yongxiang

    1999-01-01

    A probabilistic evaluating approach of design S-N curve and a reliability assessment approach of the ASME code-based evaluation are presented on the basis of Langer S-N model-based P-S-N curves. The P-S-N curves are estimated by a so-called general maximum likelihood method. This method can be applied to deal with the virtual stress amplitude-crack initial life data which have a characteristics of double random variables. Investigation of a set of the virtual stress amplitude-crack initial life (S-N) data of 1Cr18Ni9Ti austenitic stainless steel-welded joint reveals that the P-S-N curves can give a good prediction of scatter regularity of the S-N data. Probabilistic evaluation of the design S-N curve with 0.9999 survival probability has considered various uncertainties, besides of the scatter of the S-N data, to an appropriate extent. The ASME code-based evaluation with 20 reduction factor on the mean life is much more conservative than that with 2 reduction factor on the stress amplitude. Evaluation of the latter in 666.61 MPa virtual stress amplitude is equivalent to 0.999522 survival probability and in 2092.18 MPa virtual stress amplitude equivalent to 0.9999999995 survival probability. This means that the evaluation in the low loading level may be non-conservative and in contrast, too conservative in the high loading level. Cause is that the reduction factors are constants and the factors can not take into account the general observation that scatter of the N data increases with the loading level decreasing. This has indicated that it is necessary to apply the probabilistic approach to the evaluation of design S-N curve

  5. Elastic-plastic stress analysis and ASME code evaluation of a bottomhead penetration in a reactor pressure vessel

    International Nuclear Information System (INIS)

    Ranganath, S.

    1979-01-01

    Nuclear pressure vessel components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Specifically, the design must satisfy the limits on stress range and fatigue usage prescribed in NB-3200, Section III ASME Code for the various design and operating conditions for the component. The Code requirements assure that the component does not experience gross yielding and that in general, elastic shakedown occurs following cyclic loading. When elastic stress analysis is performed this can be shown by meeting the limits in the Code on Primary and Primary plus Secondary (P+Q) stress intensities. However, when the P+Q limits cannot be met and elastic Shakedown cannot be demonstrated, plastic analysis may be performed to meet the requirements of the Code. This paper describes the elastic-plastic stress analysis of a Boiling Water Reactor Vessel bottom head in-core penetration and illustrates how plastic analysis can be used in ASME Code evaluations to show Code compliance. Details of the thermal analysis, elastic-plastic stress analysis and fatigue evaluation are presented and it is shown that the in-core penetration satisfies the code requirements. 6 refs

  6. An evaluation and analysis of three dynamic watershed acidification codes (MAGIC, ETD, and ILWAS)

    Energy Technology Data Exchange (ETDEWEB)

    Jenne, E.A.; Eary, L.E.; Vail, L.W.; Girvin, D.C.; Liebetrau, A.M.; Hibler, L.F.; Miley, T.B.; Monsour, M.J.

    1989-01-01

    The US Environmental Protection Agency is currently using the dynamic watershed acidification codes MAGIC, ILWAS, and ETD to assess the potential future impact of the acidic deposition on surface water quality by simulating watershed acid neutralization processes. The reliability of forecasts made with these codes is of considerable concern. The present study evaluates the process formulations (i.e., conceptual and numerical representation of atmospheric, hydrologic geochemical and biogeochemical processes), compares their approaches to calculating acid neutralizing capacity (ANC), and estimates the relative effects (sensitivity) of perturbations in the input data on selected output variables for each code. Input data were drawn from three Adirondack (upstate New York) watersheds: Panther Lake, Clear Pond, and Woods Lake. Code calibration was performed by the developers of the codes. Conclusions focus on summarizing the adequacy of process formulations, differences in ANC simulation among codes and recommendations for further research to increase forecast reliability. 87 refs., 11 figs., 77 tabs.

  7. Evaluation of coded aperture radiation detectors using a Bayesian approach

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Kyle, E-mail: mille856@andrew.cmu.edu [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Huggins, Peter [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Labov, Simon; Nelson, Karl [Lawrence Livermore National Laboratory, Livermore, CA (United States); Dubrawski, Artur [Auton Lab, The Robotics Institute, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States)

    2016-12-11

    We investigate tradeoffs arising from the use of coded aperture gamma-ray spectrometry to detect and localize sources of harmful radiation in the presence of noisy background. Using an example application scenario of area monitoring and search, we empirically evaluate weakly supervised spectral, spatial, and hybrid spatio-spectral algorithms for scoring individual observations, and two alternative methods of fusing evidence obtained from multiple observations. Results of our experiments confirm the intuition that directional information provided by spectrometers masked with coded aperture enables gains in source localization accuracy, but at the expense of reduced probability of detection. Losses in detection performance can however be to a substantial extent reclaimed by using our new spatial and spatio-spectral scoring methods which rely on realistic assumptions regarding masking and its impact on measured photon distributions.

  8. CESARR V.2 manual: Computer code for the evaluation of surface storage of low and medium level radioactive waste

    International Nuclear Information System (INIS)

    Moya Rivera, J.A.; Bolado Lavin, R.

    1997-01-01

    CESARR (Code for the safety evaluation of low and medium level radioactive waste storage). This code was developed for the safety probabilistic evaluations in the facilities of low-and medium level radioactive waste storage

  9. GOTHIC code evaluation of alternative passive containment cooling features

    International Nuclear Information System (INIS)

    Gavrilas, M.; Todreas, E.N.; Driscoll, M.J.

    1996-01-01

    Reliance on passive cooling has become an important objective in containment design. Several reactor concepts have been set forth, which are equipped with entirely passively cooled containments. However, the problems that have to be overcome in rejecting the entire heat generated by a severe accident in a high-rating reactor (i.e. one with a rating greater than 1200 MW e ) have been found to be substantial and without obvious solutions. The GOTHIC code was verified and modified for containment cooling applications; optimal mesh sizes, computational time steps and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. The GOTHIC code was then employed to assess the effectiveness of several original heat rejection features that make it possible to cool high-rating containments. Two containment concepts were evaluated: one for a 1200 MW e new pressure tube light-water reactor, and one for a 1300 MW e pressurized-water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features has been predicted. The best-performance configurations-worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light-water reactor, and less than 0.45 MPa for the 1300 MW e pressurized-water reactor. (orig.)

  10. An accurate evaluation of the performance of asynchronous DS-CDMA systems with zero-correlation-zone coding in Rayleigh fading

    Science.gov (United States)

    Walker, Ernest; Chen, Xinjia; Cooper, Reginald L.

    2010-04-01

    An arbitrarily accurate approach is used to determine the bit-error rate (BER) performance for generalized asynchronous DS-CDMA systems, in Gaussian noise with Raleigh fading. In this paper, and the sequel, new theoretical work has been contributed which substantially enhances existing performance analysis formulations. Major contributions include: substantial computational complexity reduction, including a priori BER accuracy bounding; an analytical approach that facilitates performance evaluation for systems with arbitrary spectral spreading distributions, with non-uniform transmission delay distributions. Using prior results, augmented by these enhancements, a generalized DS-CDMA system model is constructed and used to evaluated the BER performance, in a variety of scenarios. In this paper, the generalized system modeling was used to evaluate the performance of both Walsh- Hadamard (WH) and Walsh-Hadamard-seeded zero-correlation-zone (WH-ZCZ) coding. The selection of these codes was informed by the observation that WH codes contain N spectral spreading values (0 to N - 1), one for each code sequence; while WH-ZCZ codes contain only two spectral spreading values (N/2 - 1,N/2); where N is the sequence length in chips. Since these codes span the spectral spreading range for DS-CDMA coding, by invoking an induction argument, the generalization of the system model is sufficiently supported. The results in this paper, and the sequel, support the claim that an arbitrary accurate performance analysis for DS-CDMA systems can be evaluated over the full range of binary coding, with minimal computational complexity.

  11. Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events

    International Nuclear Information System (INIS)

    Hirano, Masashi; Kohsaka, Atsuo.

    1992-02-01

    A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k I ) with the crack initiation toughness (k Ic ) and crack arrest toughness (k Ic ), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author)

  12. Evaluation of void fraction measurements from DADINE experience using RELAP4/MOD5 code

    International Nuclear Information System (INIS)

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

    1989-01-01

    The DADINE experiment measures the axial evolution of the void fraction by neutronic diffusion in two-phase flow in the wet regions of a pressurized water reactor in accident conditions. Since the theoretical/experimental confrontation is important for code evaluation, this paper presents the simulation with the RELAP4/MOD5 Code of the void fractions results obtained in the DADINE Experiment, that showed some deviation probably associated with the existing models in Code, special attention in the way of stablishing the two-phase flow and the no characterization of the differents flow regimes related with the void fractions. (author) [pt

  13. The Coding Causes of Death in HIV (CoDe) Project: initial results and evaluation of methodology

    DEFF Research Database (Denmark)

    Kowalska, Justyna D; Friis-Møller, Nina; Kirk, Ole

    2011-01-01

    The Coding Causes of Death in HIV (CoDe) Project aims to deliver a standardized method for coding the underlying cause of death in HIV-positive persons, suitable for clinical trials and epidemiologic studies.......The Coding Causes of Death in HIV (CoDe) Project aims to deliver a standardized method for coding the underlying cause of death in HIV-positive persons, suitable for clinical trials and epidemiologic studies....

  14. Evaluation of temporary non-code repairs in safety class 3 piping systems

    International Nuclear Information System (INIS)

    Godha, P.C.; Kupinski, M.; Azevedo, N.F.

    1996-01-01

    Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

  15. Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

    International Nuclear Information System (INIS)

    Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

    2012-01-01

    In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

  16. Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi; Yamamoto, Toshihiro

    1999-03-01

    In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)

  17. An algebraic approach to graph codes

    DEFF Research Database (Denmark)

    Pinero, Fernando

    This thesis consists of six chapters. The first chapter, contains a short introduction to coding theory in which we explain the coding theory concepts we use. In the second chapter, we present the required theory for evaluation codes and also give an example of some fundamental codes in coding...... theory as evaluation codes. Chapter three consists of the introduction to graph based codes, such as Tanner codes and graph codes. In Chapter four, we compute the dimension of some graph based codes with a result combining graph based codes and subfield subcodes. Moreover, some codes in chapter four...

  18. RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

    International Nuclear Information System (INIS)

    Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

    1989-03-01

    A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

  19. Comparison of FISGAS swelling and gas release predictions with experiment

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1979-01-01

    FISGAS calculations were compared to fuel swelling data from the FD1 tests and to gas release data from the FGR39 test. Late swelling and gas release predictions are satisfactory if vacancy depletion effects are added to the code. However, early swelling predictions are not satisfactory, and early gas release predictions are very poor. Explanation of these discrepancies is speculative

  20. Contrasting motivational orientation and evaluative coding accounts: On the need to differentiate the effectors of approach/avoidance responses

    Directory of Open Access Journals (Sweden)

    Julia eKozlik

    2015-05-01

    Full Text Available Several emotion theorists suggest that valenced stimuli automatically trigger motivational orientations and thereby facilitate corresponding behavior. Positive stimuli were thought to activate approach motivational circuits which in turn primed approach-related behavioral tendencies whereas negative stimuli were supposed to activate avoidance motivational circuits so that avoidance-related behavioral tendencies were primed (motivational orientation account. However, recent research suggests that typically observed affective stimulus–response compatibility phenomena might be entirely explained in terms of theories accounting for mechanisms of general action control instead of assuming motivational orientations to mediate the effects (evaluative coding account. In what follows, we explore to what extent this notion is applicable. We present literature suggesting that evaluative coding mechanisms indeed influence a wide variety of affective stimulus–response compatibility phenomena. However, the evaluative coding account does not seem to be sufficient to explain affective S–R compatibility effects. Instead, several studies provide clear evidence in favor of the motivational orientation account that seems to operate independently of evaluative coding mechanisms. Implications for theoretical developments and future research designs are discussed.

  1. Contrasting motivational orientation and evaluative coding accounts: on the need to differentiate the effectors of approach/avoidance responses.

    Science.gov (United States)

    Kozlik, Julia; Neumann, Roland; Lozo, Ljubica

    2015-01-01

    Several emotion theorists suggest that valenced stimuli automatically trigger motivational orientations and thereby facilitate corresponding behavior. Positive stimuli were thought to activate approach motivational circuits which in turn primed approach-related behavioral tendencies whereas negative stimuli were supposed to activate avoidance motivational circuits so that avoidance-related behavioral tendencies were primed (motivational orientation account). However, recent research suggests that typically observed affective stimulus-response compatibility phenomena might be entirely explained in terms of theories accounting for mechanisms of general action control instead of assuming motivational orientations to mediate the effects (evaluative coding account). In what follows, we explore to what extent this notion is applicable. We present literature suggesting that evaluative coding mechanisms indeed influence a wide variety of affective stimulus-response compatibility phenomena. However, the evaluative coding account does not seem to be sufficient to explain affective S-R compatibility effects. Instead, several studies provide clear evidence in favor of the motivational orientation account that seems to operate independently of evaluative coding mechanisms. Implications for theoretical developments and future research designs are discussed.

  2. Cracking the code: the accuracy of coding shoulder procedures and the repercussions.

    Science.gov (United States)

    Clement, N D; Murray, I R; Nie, Y X; McBirnie, J M

    2013-05-01

    Coding of patients' diagnosis and surgical procedures is subject to error levels of up to 40% with consequences on distribution of resources and financial recompense. Our aim was to explore and address reasons behind coding errors of shoulder diagnosis and surgical procedures and to evaluate a potential solution. A retrospective review of 100 patients who had undergone surgery was carried out. Coding errors were identified and the reasons explored. A coding proforma was designed to address these errors and was prospectively evaluated for 100 patients. The financial implications were also considered. Retrospective analysis revealed the correct primary diagnosis was assigned in 54 patients (54%) had an entirely correct diagnosis, and only 7 (7%) patients had a correct procedure code assigned. Coders identified indistinct clinical notes and poor clarity of procedure codes as reasons for errors. The proforma was significantly more likely to assign the correct diagnosis (odds ratio 18.2, p code (odds ratio 310.0, p coding department. High error levels for coding are due to misinterpretation of notes and ambiguity of procedure codes. This can be addressed by allowing surgeons to assign the diagnosis and procedure using a simplified list that is passed directly to coding.

  3. GOTHIC code evaluation of alternative passive containment cooling features

    International Nuclear Information System (INIS)

    Gavrilas, M.; Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1994-01-01

    The GOTHIC code was employed to assess the effectiveness of several original heat rejection features that make it possible to cool large rating containments. The code was first verified and modified for specific containment cooling applications; optimal mesh sizes, computational time steps, and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. GOTHIC was then used to obtain performance predictions for two containment concepts: a 1200 MW e new pressure tube light water reactor, and a 1300 MW e pressurized water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features have been predicted. For the 1200 MW e pressure tube light water reactor, the evaluated pressure-limiting features are: a large water pool connected to the calandria, large containment free volume and an air-convection annulus. For the 1300 MW e pressurized water reactor, an external moat, an internal water pool, and an air-convection annulus were evaluated. The performance of the proposed containment configurations is dependent on the extent of thermal stratification inside the containment. The best-performance configurations/worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light water reactor, and less than 0.45 MPa for the 1300 MW e pressurized water reactor. The low peak pressure predicted for the 1200 MW e pressure tube light water reactor can be in part attributed to its relatively large free volume, while the relatively high peak pressure predicted for the 1300 MW e pressurized water reactor can be attributed to its relatively small free volume (i.e., the size used was that of a pressurized water reactor containment designed with active heat removal features). (author)

  4. WAMCUT, a computer code for fault tree evaluation. Final report

    International Nuclear Information System (INIS)

    Erdmann, R.C.

    1978-06-01

    WAMCUT is a code in the WAM family which produces the minimum cut sets (MCS) for a given fault tree. The MCS are useful as they provide a qualitative evaluation of a system, as well as providing a means of determining the probability distribution function for the top of the tree. The program is very efficient and will produce all the MCS in a very short computer time span. 22 figures, 4 tables

  5. Investigation of knowledge structure of nuclear data evaluation code

    International Nuclear Information System (INIS)

    Uenaka, Junji; Kambayashi, Shaw

    1988-08-01

    In this report, investigation results of knowledge structure in a nuclear data evaluation code are described. This investigation is related to the natural language processing and the knowledge base in the research theme of Human Acts Simulation Program (HASP) begun at the Computing Center of JAERI in 1987. By using a machine translation system, an attempt has been made to extract a deep knowledge from Japanese sentences which are equivalent to a FORTRAN program CASTHY for nuclear data evaluation. With the knowledge extraction method used by the authors, the verification of knowledge is more difficult than that of the prototyping method in an ordinary AI technique. In the early stage of building up a knowledge base system, it seems effective to extract and examine knowledge fragments of limited objects. (author)

  6. Evaluation of hydrogen production system coupling with HTTR using dynamic analysis code

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Hayashi, Koji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency (JAEA) was entrusted 'Development of Nuclear Heat Utilization Technology' by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well. (author)

  7. Kinetic parameters evaluation of PWRs using static cell and core calculation codes

    International Nuclear Information System (INIS)

    Jahanbin, Ali; Malmir, Hessam

    2012-01-01

    Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((β eff ) i )/((β eff ) core ) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.

  8. Implementation and Performance Evaluation of Distributed Cloud Storage Solutions using Random Linear Network Coding

    DEFF Research Database (Denmark)

    Fitzek, Frank; Toth, Tamas; Szabados, Áron

    2014-01-01

    This paper advocates the use of random linear network coding for storage in distributed clouds in order to reduce storage and traffic costs in dynamic settings, i.e. when adding and removing numerous storage devices/clouds on-the-fly and when the number of reachable clouds is limited. We introduce...... various network coding approaches that trade-off reliability, storage and traffic costs, and system complexity relying on probabilistic recoding for cloud regeneration. We compare these approaches with other approaches based on data replication and Reed-Solomon codes. A simulator has been developed...... to carry out a thorough performance evaluation of the various approaches when relying on different system settings, e.g., finite fields, and network/storage conditions, e.g., storage space used per cloud, limited network use, and limited recoding capabilities. In contrast to standard coding approaches, our...

  9. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  10. Sildenafil citrate (Viagra) enhances vasodilatation in fetal growth restriction.

    Science.gov (United States)

    Wareing, Mark; Myers, Jenny E; O'Hara, Maureen; Baker, Philip N

    2005-05-01

    Fetal growth restriction (FGR) affects up to 8% of all pregnancies and has massive short-term (increased fetal morbidity and mortality) and long-term (increased incidence of cardiovascular disease in adulthood) health implications. Doppler waveform analysis of pregnancies complicated by FGR suggests compromised uteroplacental circulation and placental hypoperfusion. Our aim was to determine whether myometrial small artery function was aberrant in FGR and to assess whether sildenafil citrate could improve vasodilatation in FGR pregnancies. Small arteries dissected from myometrial biopsies obtained at cesarean section from normal pregnant women (n = 27) or women whose pregnancies were complicated by FGR (n = 12) were mounted on wire myographs. Vessels were constricted (with arginine vasopressin or U46619) and relaxed (with bradykinin) before and after incubation with a phosphodiesterase-5 inhibitor, sildenafil citrate. We demonstrated increased myometrial small artery vasoconstriction and decreased endothelium-dependent vasodilatation in vessels from women whose pregnancies were complicated by FGR. Sildenafil citrate significantly reduced vasoconstriction and significantly improved relaxation of FGR small arteries. We conclude that sildenafil citrate improves endothelial function of myometrial vessels from women whose pregnancies are complicated by intrauterine growth restriction. Sildenafil citrate may offer a potential therapeutic strategy to improve uteroplacental blood flow in FGR pregnancies.

  11. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  12. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y. [Westinghouse Electric Company LLC, Pittsburgh (United States)

    2011-07-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  13. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    International Nuclear Information System (INIS)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y.

    2011-01-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  14. The missing evaluation codes from order domain theory

    DEFF Research Database (Denmark)

    Andersen, Henning Ejnar; Geil, Olav

    The Feng-Rao bound gives a lower bound on the minimum distance of codes defined by means of their parity check matrices. From the Feng-Rao bound it is clear how to improve a large family of codes by leaving out certain rows in their parity check matrices. In this paper we derive a simple lower...... generalized Hamming weight. We interpret our methods into the setting of order domain theory. In this way we fill in an obvious gap in the theory of order domains. The improved codes from the present paper are not in general equal to the Feng-Rao improved codes but the constructions are very much related....

  15. External dose-rate conversion factors of radionuclides for air submersion, ground surface contamination and water immersion based on the new ICRP dosimetric setting.

    Science.gov (United States)

    Yoo, Song Jae; Jang, Han-Ki; Lee, Jai-Ki; Noh, Siwan; Cho, Gyuseong

    2013-01-01

    For the assessment of external doses due to contaminated environment, the dose-rate conversion factors (DCFs) prescribed in Federal Guidance Report 12 (FGR 12) and FGR 13 have been widely used. Recently, there were significant changes in dosimetric models and parameters, which include the use of the Reference Male and Female Phantoms and the revised tissue weighting factors, as well as the updated decay data of radionuclides. In this study, the DCFs for effective and equivalent doses were calculated for three exposure settings: skyshine, groundshine and water immersion. Doses to the Reference Phantoms were calculated by Monte Carlo simulations with the MCNPX 2.7.0 radiation transport code for 26 mono-energy photons between 0.01 and 10 MeV. The transport calculations were performed for the source volume within the cut-off distances practically contributing to the dose rates, which were determined by a simplified calculation model. For small tissues for which the reduction of variances are difficult, the equivalent dose ratios to a larger tissue (with lower statistical errors) nearby were employed to make the calculation efficient. Empirical response functions relating photon energies, and the organ equivalent doses or the effective doses were then derived by the use of cubic-spline fitting of the resulting doses for 26 energy points. The DCFs for all radionuclides considered important were evaluated by combining the photon emission data of the radionuclide and the empirical response functions. Finally, contributions of accompanied beta particles to the skin equivalent doses and the effective doses were calculated separately and added to the DCFs. For radionuclides considered in this study, the new DCFs for the three exposure settings were within ±10 % when compared with DCFs in FGR 13.

  16. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantly affects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs

  17. Preliminary analysis of in-reactor behavior of three MOX fuel rods in the halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong; Joo, Hyung Kook

    1999-09-01

    Preliminary analysis of in-reactor thermal performance for three MOX fuel rods that are going to be irradiated in the Halden reactor from the first quarter of the year 2000 have been conducted by using the computer code COSMOS. Using the assumption that microstructure of MOX fuel fabricated by SBR and dry milling method is the same, parametric studies have been carried out considering four kinds of uncertainties, which are thermal conductivity, linear power, manufacturing parameters, and model constant, to investigate the effect of each of uncertainty on in-reactor behavior. It is found that the uncertainty of model constants for FGR has a greatest impact of the all because the amount of gas released to the gap is one of the parameters that dominantlyaffects the gap conductance. The parametric analysis shows that, tn the case of MOX-1, calculational results vary widely depending on the choice of model constants for FGR. Therefore, the model constants for FGR for the present test need to be established through the measured fuel centerline temperature, rod internal pressure, stack length if any, and finally thermal conductivity derived from measured data during irradiation. On the other hand, the difference in thermal performance of MOX-3 resulting from the choice of FGR model constants is not so large as that for MOX-1. This might arise, since the temperature of the MOX-3 is high, the capacity of grain boundaries to retain gas atoms is not sufficient enough to accommodate the large amount of gas atoms reaching the grain boundaries through diffusion. (Author). 20 refs., 7 tabs., 47 figs.

  18. Diagnostic Coding for Epilepsy.

    Science.gov (United States)

    Williams, Korwyn; Nuwer, Marc R; Buchhalter, Jeffrey R

    2016-02-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  19. Coding of Neuroinfectious Diseases.

    Science.gov (United States)

    Barkley, Gregory L

    2015-12-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  20. Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)

    2015-10-15

    The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.

  1. Effect of behavior training on learning and memory of young rats with fetal growth restriction

    Institute of Scientific and Technical Information of China (English)

    Li Xuelan; Gou Wenli; Huang Pu; Li Chunfang; Sun Yunping

    2008-01-01

    Objective: To investigate the effect of behavior training on the learning and memory of young rats with fetal growth restriction (FGR). Methods: The model of FGR was established by passive smoking method to pregnant rats.The new-born rats were divided into FGR group and normal group, and then randomly subdivided into trained and untrained group respectively. Morris water maze behavior training was performed on postnatal months 2 and 4, then learning and memory abilities of young rats were measured by dark-avoidance testing and step-down testing. Results: In the dark-avoidance and step-down testing, the young rats' performance of FGR group was worse than that of control group, and the trained group was better than the untrained group significantly. Conclusion: FGR young rats have descended learning and memory abilities. Behavior training could improve the young rats' learning and memory abilities, especially for the FGR young rats.

  2. [Effect of antepartum taurine supplementation in regulating the activity of Rho family factors and promoting the proliferation of neural stem cells in neonatal rats with fetal growth restriction].

    Science.gov (United States)

    Li, Xiang-Wen; Li, Fang; Liu, Jing; Wang, Yan; Fu, Wei

    2016-11-01

    To study the possible effect of antepartum taurine supplementation in regulating the activity of Rho family factors and promoting the proliferation of neural stem cells in neonatal rats with fetal growth restriction (FGR), and to provide a basis for antepartum taurine supplementation to promote brain development in children with FGR. A total of 24 pregnant Sprague-Dawley rats were randomly divided into three groups: control, FGR, and taurine (n=8 each ). A rat model of FGR was established by food restriction throughout pregnancy. RT-PCR, immunohistochemistry, and Western blot were used to measure the expression of the specific intracellular markers for neural stem cells fatty acid binding protein 7 (FABP7), Rho-associated coiled-coil containing protein kinase 2 (ROCK2), ras homolog gene family, member A (RhoA), and Ras-related C3 botulinum toxin substrate (Rac). The FGR group had significantly lower OD value of FABP7-positive cells and mRNA and protein expression of FABP7 than the control group, and the taurine group had significantly higher OD value of FABP7-positive cells and mRNA and protein expression of FABP7 than the FGR group (Ptaurine group had significantly higher mRNA expression of RhoA and ROCK2 than the control group and significantly lower expression than the FGR group (Ptaurine group had significantly higher mRNA expression of Rac than the FGR and control groups (Ptaurine group had significantly lower protein expression of RhoA and ROCK2 than the FGR group (Ptaurine supplementation can promote the proliferation of neural stem cells in rats with FGR, and its mechanism may be related to the regulation of the activity of Rho family factors.

  3. Computer-modeling codes to improve exploration nuclear-logging methods. National Uranium Resource Evaluation

    International Nuclear Information System (INIS)

    Wilson, R.D.; Price, R.K.; Kosanke, K.L.

    1983-03-01

    As part of the Department of Energy's National Uranium Resource Evaluation (NURE) project's Technology Development effort, a number of computer codes and accompanying data bases were assembled for use in modeling responses of nuclear borehole logging Sondes. The logging methods include fission neutron, active and passive gamma-ray, and gamma-gamma. These CDC-compatible computer codes and data bases are available on magnetic tape from the DOE Technical Library at its Grand Junction Area Office. Some of the computer codes are standard radiation-transport programs that have been available to the radiation shielding community for several years. Other codes were specifically written to model the response of borehole radiation detectors or are specialized borehole modeling versions of existing Monte Carlo transport programs. Results from several radiation modeling studies are available as two large data bases (neutron and gamma-ray). These data bases are accompanied by appropriate processing programs that permit the user to model a wide range of borehole and formation-parameter combinations for fission-neutron, neutron-, activation and gamma-gamma logs. The first part of this report consists of a brief abstract for each code or data base. The abstract gives the code name and title, short description, auxiliary requirements, typical running time (CDC 6600), and a list of references. The next section gives format specifications and/or directory for the tapes. The final section of the report presents listings for programs used to convert data bases between machine floating-point and EBCDIC

  4. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    International Nuclear Information System (INIS)

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files

  5. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.

  6. Error floor behavior study of LDPC codes for concatenated codes design

    Science.gov (United States)

    Chen, Weigang; Yin, Liuguo; Lu, Jianhua

    2007-11-01

    Error floor behavior of low-density parity-check (LDPC) codes using quantized decoding algorithms is statistically studied with experimental results on a hardware evaluation platform. The results present the distribution of the residual errors after decoding failure and reveal that the number of residual error bits in a codeword is usually very small using quantized sum-product (SP) algorithm. Therefore, LDPC code may serve as the inner code in a concatenated coding system with a high code rate outer code and thus an ultra low error floor can be achieved. This conclusion is also verified by the experimental results.

  7. Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohya, Kaoru; Inai, Kensuke [Univ. of Tokushima, Institute of Technology and Science, Tokushima, Tokushima (Japan); Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan); Hatayama, Akiyoshi; Toma, Mitsunori [Keio Univ., Faculty of Science and Technology, Yokohama, Kanagawa (Japan); Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji [National Inst. for Fusion Science, Toki, Gifu (Japan); Tanaka, Yasunori [Kanazawa Univ., College of Science and Engineering, Kanazawa, Ishikawa (Japan); Ono, Tadayoshi; Muramoto, Tetsuya [Okayama Univ. of Science, Faculty of Informatics, Okayama, Okayama (Japan); Kenmotsu, Takahiro [Doshisha Univ., Faculty of Life and Medical Science, Kiyotanabe, Kyoto (Japan)

    2009-10-15

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

  8. Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

    International Nuclear Information System (INIS)

    Ohya, Kaoru; Inai, Kensuke; Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo; Hatayama, Akiyoshi; Toma, Mitsunori; Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji; Tanaka, Yasunori; Ono, Tadayoshi; Muramoto, Tetsuya; Kenmotsu, Takahiro

    2009-01-01

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

  9. Catch-up growth in children born growth restricted to mothers with hypertensive disorders of pregnancy

    NARCIS (Netherlands)

    Beukers, Fenny; Cranendonk, Anneke; de Vries, Johanna I. P.; Wolf, Hans; Lafeber, Harry N.; Vriesendorp, Hester C.; Ganzevoort, Wessel; van Wassenaer-Leemhuis, Aleid G.

    2013-01-01

    In preterm hypertensive disorders of pregnancy, fetal growth restriction (FGR) occurs frequently. The timing and severity of FGR impacts childhood growth and is associated with metabolic changes later in life. To examine growth and the impact of FGR in early childhood. Prospective cohort study.

  10. Nonadiabatic Dynamics May Be Probed through Electronic Coherence in Time-Resolved Photoelectron Spectroscopy.

    Science.gov (United States)

    Bennett, Kochise; Kowalewski, Markus; Mukamel, Shaul

    2016-02-09

    We present a hierarchy of Fermi golden rules (FGRs) that incorporate strongly coupled electronic/nuclear dynamics in time-resolved photoelectron spectroscopy (TRPES) signals at different levels of theory. Expansion in the joint electronic and nuclear eigenbasis yields the numerically most challenging exact FGR (eFGR). The quasistatic Fermi Golden Rule (qsFGR) neglects nuclear motion during the photoionization process but takes into account electronic coherences as well as populations initially present in the pumped matter as well as those generated internally by coupling between electronic surfaces. The standard semiclassical Fermi Golden Rule (scFGR) neglects the electronic coherences and the nuclear kinetic energy during the ionizing pulse altogether, yielding the classical Condon approximation. The coherence contributions depend on the phase-profile of the ionizing field, allowing coherent control of TRPES signals. The photoelectron spectrum from model systems is simulated using these three levels of theory. The eFGR and the qsFGR show temporal oscillations originating from the electronic or vibrational coherences generated as the nuclear wave packet traverses a conical intersection. These oscillations, which are missed by the scFGR, directly reveal the time-evolving splitting between electronic states of the neutral molecule in the curve-crossing regime.

  11. Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

  12. Coding in Muscle Disease.

    Science.gov (United States)

    Jones, Lyell K; Ney, John P

    2016-12-01

    Accurate coding is critically important for clinical practice and research. Ongoing changes to diagnostic and billing codes require the clinician to stay abreast of coding updates. Payment for health care services, data sets for health services research, and reporting for medical quality improvement all require accurate administrative coding. This article provides an overview of administrative coding for patients with muscle disease and includes a case-based review of diagnostic and Evaluation and Management (E/M) coding principles in patients with myopathy. Procedural coding for electrodiagnostic studies and neuromuscular ultrasound is also reviewed.

  13. On the use of the HOTSPOT code for evaluating accidents involving radioactive materials

    International Nuclear Information System (INIS)

    Sattinger, D.; Sarussi, R.; Tzarfati, Y.; Levinson, S.; Tshuva, A.

    2004-01-01

    The HOTSPOT Health Physics code was created by LLNL in order to provide Health Physics personnel with a fast, field portable calculation tool for evaluating accidents involving radioactive materials. The HOTSPOT code is a first order approximation of the radiation effects associated with the atmospheric release of radioactive materials. HOTSPOT programs are reasonably accurate for a timely initial assessment. More importantly, HOTSPOT code produce a consistent output for the same input assumptions, and minimize the probability of errors associated with reading a graph incorrectly. Four general programs, Plume, Explosion, Fire, and Resuspension, calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel or fire, or an area contamination event. Additional programs estimate the dose commitment from inhalation of any one of the radionuclides listed in the database of radionuclides, calibrate a radiation survey instrument for ground survey measurements, and screening of alpha emitters in the Lung. We believe that the HOTSPOT code is extremely valuable in providing reasonable and reliable guidance for a diversity of application. For example, we demonstrate the release of 241 Am(20Ci) to the atmosphere

  14. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  15. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  16. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in

  17. RSE-M code progress in the field of examination evaluation and flaw acceptance criteria

    International Nuclear Information System (INIS)

    Barthelet, B.; Le Delliou, P.; Heliot, J.; Faidy, C.; Drubay, B.

    1995-01-01

    The RSE-M Code provides rules and requirements for in service inspection of light water cooled nuclear power plants. The code first edition was established by EDF and published in 1990 by AFCEN. In 1992, a second RSE-M project was launched by EDF and FRAMATOME with the objective to address a 1995 edition more completed considering the needs of owners, users, manufacturers and inspectors. This paper focuses on evaluation of examination results and presents the work done in the field of flaw acceptance criteria over the last three years. (author). 5 refs., 3 figs

  18. Fetal Growth Restriction with Brain Sparing: Neurocognitive and Behavioral Outcomes at 12 Years of Age

    NARCIS (Netherlands)

    Beukers, Fenny; Aarnoudse-Moens, Cornelieke S. H.; van Weissenbruch, Mirjam M.; Ganzevoort, Wessel; van Goudoever, Johannes B.; van Wassenaer-Leemhuis, Aleid G.

    2017-01-01

    Objective To study neurocognitive functions and behavior in children with a history of fetal growth restriction (FGR) with brain sparing. We hypothesized that children with FGR would have poorer outcomes on these domains. Study design Subjects were 12-year-old children with a history of FGR born to

  19. From head to heart; : the effects of fetal growth restriction and preterm birth on the cerebral and systemic circulation

    NARCIS (Netherlands)

    Cohen, Emily

    2017-01-01

    Fetal growth restriction (FGR) is the condition where a fetus does not grow according to its genetic growth potential. It is estimated that 3-7% of pregnancies are complicated by FGR. FGR has been associated with many adverse outcomes, including an increased risk of perinatal and neonatal morbidity

  20. What Do We Know about Risk Factors for Fetal Growth Restriction in Africa at the Time of Sustainable Development Goals? A Scoping Review.

    Science.gov (United States)

    Accrombessi, Manfred; Zeitlin, Jennifer; Massougbodji, Achille; Cot, Michel; Briand, Valérie

    2018-03-01

    The reduction in the under-5 year mortality rate to at least as low as 25 per 1000 livebirths by 2030 has been implemented as one of the new Sustainable Development Goals. Fetal growth restriction (FGR) is one of the most important determinants of infant mortality in developing countries. In this review, we assess the extent of the literature and summarize its findings on the main preventable factors of FGR in Africa. A scoping review was conducted using the Arksey and O'Malley framework. Five bibliographic databases and grey literature were used to identify studies assessing at least one risk factor for FGR. Aggregate risk estimates for the main factors associated with FGR were calculated. Forty-five of a total of 671 articles were selected for the review. The prevalence of FGR varied between 2.6 and 59.2% according to both the African region and the definition of FGR. The main preventable factors reported were a low maternal nutritional status (aggrerate odds ratio [OR]: 2.28, 95% confidence interval [CI] 1.59, 3.25), HIV infection (aOR 1.86, 95% CI 1.38, 2.50), malaria (aOR 1.95, 95% CI 1.04, 3.66), and gestational hypertension (aOR 2.61, 95% CI 2.42, 2.82). FGR is, to a large extent, preventable through existing efficacious interventions dedicated to malaria, HIV and nutrition. Further studies are still needed to assess the influence of risk factors most commonly documented in high-income countries. Improving research on FGR in Africa requires a consensual and standardized definition of FGR-for a higher comparability-between studies and settings. © 2017 John Wiley & Sons Ltd.

  1. Evaluation of Thermal Load to the Lower Head Vessel Using the ASTEC Computer Code

    International Nuclear Information System (INIS)

    Park, Raejoon; Ahn, Kwangil

    2013-01-01

    The thermal load from the corium to the lower head vessel in the APR (Advanced Power reactor) 1400 during a small break loss of coolant accident (SBLOCA) without a safety injection (SI) has been evaluated using the ASTEC (Accident Source Term Evaluation Code) computer code, which has been developed as a part of the EU (European Union)-SARNET (Severe Accident Research NET work) program. The ASTEC results predict that the reactor vessel did not fail by using an ERVC, in spite of the large melting of the reactor vessel wall in a two-layer formation case of the SBLOCA in the APR1400. The outer surface conditions of the temperature and heat transfer coefficient are not effective on the vessel geometry change, which are preliminary results. A more detailed analysis of the main parameter effects on the corium behavior in the lower plenum is necessary to evaluate the IVR-ERVC in the APR1400, in particular, for a three-layer formation of the TLFW. Comparisons of the present results with others are necessary to verify and apply them to the actual IVR-ERVC evaluation in the APR1400

  2. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  3. Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

    1996-12-01

    The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

  4. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

    2016-11-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  5. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    International Nuclear Information System (INIS)

    Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

    2016-01-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  6. A computer code SPHINCS for sodium fire safety evaluation

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2000-01-01

    A computer code SPHINCS solves coupled phenomena of thermal-hydraulics and sodium fire based on a multi-zone model. It deals with arbitrary number of rooms each of which is connected mutually by doorway and penetrations. With regard to the combustion phenomena, flame sheet model and liquid droplet combustion model are used for pool and spray fire, respectively, with the chemical equilibrium model using Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermal-hydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The author analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS has been validated with regard to the pool combustion phenomena. Further research needs are identified for the pool spreading modeling considering thermal deformation of liner and measurement of pool fluidity property of a mixture of liquid sodium and reaction products. SPHINCS code is to be used mainly in the safety evaluation of the consequence of sodium fire accident of liquid metal cooled fast reactor. (author)

  7. Quantitative code accuracy evaluation of ISP33

    Energy Technology Data Exchange (ETDEWEB)

    Kalli, H.; Miwrrin, A. [Lappeenranta Univ. of Technology (Finland); Purhonen, H. [VTT Energy, Lappeenranta (Finland)] [and others

    1995-09-01

    Aiming at quantifying code accuracy, a methodology based on the Fast Fourier Transform has been developed at the University of Pisa, Italy. The paper deals with a short presentation of the methodology and its application to pre-test and post-test calculations submitted to the International Standard Problem ISP33. This was a double-blind natural circulation exercise with a stepwise reduced primary coolant inventory, performed in PACTEL facility in Finland. PACTEL is a 1/305 volumetrically scaled, full-height simulator of the Russian type VVER-440 pressurized water reactor, with horizontal steam generators and loop seals in both cold and hot legs. Fifteen foreign organizations participated in ISP33, with 21 blind calculations and 20 post-test calculations, altogether 10 different thermal hydraulic codes and code versions were used. The results of the application of the methodology to nine selected measured quantities are summarized.

  8. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  9. Trisomy 9 Mosaicism Diagnosed In Utero

    Directory of Open Access Journals (Sweden)

    Hironori Takahashi

    2010-01-01

    Full Text Available We present three cases of trisomy 9 mosaicism diagnosed by amniocentesis with ongoing pregnancies after referral to our center due to fetal abnormalities. Two cases were associated with severe fetal growth restriction (FGR, each of which resulted in an intrauterine fetal demise (IUFD in the third trimester. The other case involved mild FGR with a congenital diaphragmatic hernia and resulted in a live birth with severe development delay. A major prenatal finding of trisomy 9 mosaicism is FGR. Fetuses with trisomy 9 mosaicism can rarely survive in the case of severe FGR.

  10. User's manual of a computer code for seismic hazard evaluation for assessing the threat to a facility by fault model. SHEAT-FM

    International Nuclear Information System (INIS)

    Sugino, Hideharu; Onizawa, Kunio; Suzuki, Masahide

    2005-09-01

    To establish the reliability evaluation method for aged structural component, we developed a probabilistic seismic hazard evaluation code SHEAT-FM (Seismic Hazard Evaluation for Assessing the Threat to a facility site - Fault Model) using a seismic motion prediction method based on fault model. In order to improve the seismic hazard evaluation, this code takes the latest knowledge in the field of earthquake engineering into account. For example, the code involves a group delay time of observed records and an update process model of active fault. This report describes the user's guide of SHEAT-FM, including the outline of the seismic hazard evaluation, specification of input data, sample problem for a model site, system information and execution method. (author)

  11. ANT: Software for Generating and Evaluating Degenerate Codons for Natural and Expanded Genetic Codes.

    Science.gov (United States)

    Engqvist, Martin K M; Nielsen, Jens

    2015-08-21

    The Ambiguous Nucleotide Tool (ANT) is a desktop application that generates and evaluates degenerate codons. Degenerate codons are used to represent DNA positions that have multiple possible nucleotide alternatives. This is useful for protein engineering and directed evolution, where primers specified with degenerate codons are used as a basis for generating libraries of protein sequences. ANT is intuitive and can be used in a graphical user interface or by interacting with the code through a defined application programming interface. ANT comes with full support for nonstandard, user-defined, or expanded genetic codes (translation tables), which is important because synthetic biology is being applied to an ever widening range of natural and engineered organisms. The Python source code for ANT is freely distributed so that it may be used without restriction, modified, and incorporated in other software or custom data pipelines.

  12. FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

    Energy Technology Data Exchange (ETDEWEB)

    Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

  13. Critical Care Coding for Neurologists.

    Science.gov (United States)

    Nuwer, Marc R; Vespa, Paul M

    2015-10-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  14. [Intelligence level and structure in school age children with fetal growth restriction].

    Science.gov (United States)

    Ma, Jian; Ma, Hong-Wei; Tian, Xiao-Bo; Liu, Fang

    2009-10-01

    To study the intelligence level and structure in school age children with fetal growth restriction (FGR). The intelligence levels were tested by the Wechsler Children Scales of Intelligence (C-WISC) in 54 children with FGR and in 84 normal children. The full intelligence quotient (FIQ), verbal IQ (VIQ) and performance IQ (PIQ) in the FGR group were 105.9+/-10.3, 112.4+/-11.2 and 97.1+/-10.6 respectively, and they all were in a normal range. But the PIQ was significantly lower than that in the control group (104.8+/-10.5; pintelligence level of children with FGR is normal, but there are imbalances in the intelligence structure and dysfunctions in performance ability related to right cerebral hemisphere. Performance trainings should be done from the infancy in children with FGR.

  15. Catch-up growth following fetal growth restriction promotes rapid restoration of fat mass but without metabolic consequences at one year of age.

    Directory of Open Access Journals (Sweden)

    Jacques Beltrand

    Full Text Available BACKGROUND: Fetal growth restriction (FGR followed by rapid weight gain during early life has been suggested to be the initial sequence promoting central adiposity and insulin resistance. However, the link between fetal and early postnatal growth and the associated anthropometric and metabolic changes have been poorly studied. METHODOLOGY/PRINCIPAL FINDINGS: Over the first year of post-natal life, changes in body mass index, skinfold thickness and hormonal concentrations were prospectively monitored in 94 infants in whom the fetal growth velocity had previously been measured using a repeated standardized procedure of ultrasound fetal measurements. 45 infants, thinner at birth, had experienced previous FGR (FGR+ regardless of birth weight. Growth pattern in the first four months of life was characterized by greater change in BMI z-score in FGR+ (+1.26+/-1.2 vs +0.58 +/-1.17 SD in FGR- resulting in the restoration of BMI and of fat mass to values similar to FGR-, independently of caloric intakes. Growth velocity after 4 months was similar and BMI z-score and fat mass remained similar at 12 months of age. At both time-points, fetal growth velocity was an independent predictor of fat mass in FGR+. At one year, fasting insulin levels were not different but leptin was significantly higher in the FGR+ (4.43+/-1.41 vs 2.63+/-1 ng/ml in FGR-. CONCLUSION: Early catch-up growth is related to the fetal growth pattern itself, irrespective of birth weight, and is associated with higher insulin sensitivity and lower leptin levels after birth. Catch-up growth promotes the restoration of body size and fat stores without detrimental consequences at one year of age on body composition or metabolic profile. The higher leptin concentration at one year may reflect a positive energy balance in children who previously faced fetal growth restriction.

  16. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    International Nuclear Information System (INIS)

    Tso, C.F.; Hueggenberg, R.

    2004-01-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work

  17. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Tso, C.F. [Arup (United Kingdom); Hueggenberg, R. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work.

  18. Training and support to improve ICD coding quality: A controlled before-and-after impact evaluation.

    Science.gov (United States)

    Dyers, Robin; Ward, Grant; Du Plooy, Shane; Fourie, Stephanus; Evans, Juliet; Mahomed, Hassan

    2017-05-24

    The proposed National Health Insurance policy for South Africa (SA) requires hospitals to maintain high-quality International Statistical Classification of Diseases (ICD) codes for patient records. While considerable strides had been made to improve ICD coding coverage by digitising the discharge process in the Western Cape Province, further intervention was required to improve data quality. The aim of this controlled before-and-after study was to evaluate the impact of a clinician training and support initiative to improve ICD coding quality. To compare ICD coding quality between two central hospitals in the Western Cape before and after the implementation of a training and support initiative for clinicians at one of the sites. The difference in differences in data quality between the intervention site and the control site was calculated. Multiple logistic regression was also used to determine the odds of data quality improvement after the intervention and to adjust for potential differences between the groups. The intervention had a positive impact of 38.0% on ICD coding completeness over and above changes that occurred at the control site. Relative to the baseline, patient records at the intervention site had a 6.6 (95% confidence interval 3.5 - 16.2) adjusted odds ratio of having a complete set of ICD codes for an admission episode after the introduction of the training and support package. The findings on impact on ICD coding accuracy were not significant. There is sufficient pragmatic evidence that a training and support package will have a considerable positive impact on ICD coding completeness in the SA setting.

  19. Computer code structure for evaluation of fire protection measures and fighting capability at nuclear plants

    International Nuclear Information System (INIS)

    Anton, V.

    1997-01-01

    In this work a computer code structure for Fire Protection Measures (FPM) and Fire Fighting Capability (FFC) at Nuclear Power Plants (NPP) is presented. It allows to evaluate the category (satisfactory (s), needs for further evaluation (n), unsatisfactory (u)) to which belongs the given NPP for a self-control in view of an IAEA inspection. This possibility of a self assessment resulted from IAEA documents. Our approach is based on international experience gained in this field and stated in IAEA recommendations. As an illustration we used the FORTRAN programming language statement to make clear the structure of the computer code for the problem taken into account. This computer programme can be conceived so that some literal message in English and Romanian languages be displayed beside the percentage assessments. (author)

  20. Mixed-oxide (MOX) fuel performance benchmark. Summary of the results for the PRIMO MOX rod BD8

    International Nuclear Information System (INIS)

    Ott, L.J.; Sartori, E.; Costa, A.; ); Sobolev, V.; Lee, B-H.; Alekseev, P.N.; Shestopalov, A.A.; Mikityuk, K.O.; Fomichenko, P.A.; Shatrova, L.P.; Medvedev, A.V.; Bogatyr, S.M.; Khvostov, G.A.; Kuznetsov, V.I.; Stoenescu, R.; Chatwin, C.P.

    2009-01-01

    The OECD/NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, nuclear fuel performance, and fuel cycle issues related to the disposition of weapons-grade plutonium as MOX fuel. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close cooperation with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A major part of these activities includes benchmark studies. This report describes the results of the PRIMO rod BD8 benchmark exercise, the second benchmark by the TFRPD relative to MOX fuel behaviour. The corresponding PRIMO experimental data have been released, compiled and reviewed for the International Fuel Performance Experiments (IFPE) database. The observed ranges (as noted in the text) in the predicted thermal and FGR responses are reasonable given the variety and combination of thermal conductivity and FGR models employed by the benchmark participants with their respective fuel performance codes

  1. Assessment of the Effects on PCT Evaluation of Enhanced Fuel Model Facilitated by Coupling the MARS Code with the FRAPTRAN Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyong Chol; Lee, Young Jin; Han, Sam Hee [NSE Technology Inc., Daejeon (Korea, Republic of)

    2016-10-15

    The principal objectives of the two safety criteria, peak cladding temperature (PCT) and total oxidation limits, are to ensure that the fuel rod claddings remain sufficiently ductile so that they do not crack and fragment during a LOCA. Another important purpose of the PCT limit is to ensure that the fuel cladding does not enter the regime of runaway oxidation and uncontrollable heat-up. However, even when the PCT limit is satisfied, it is known that cladding failures may still occur in a certain percentage of the fuel rods during a LOCA. This is largely because a 100% fuel failure is assumed for the radiological consequence analysis in the US regulatory practices. In this study, we analyze the effects of cladding failure and other fuel model features on PCT during a LOCA using the MARS-FRAPTRAN coupled code. MARS code has been coupled with FRAPTRAN code to extend fuel modeling capability. The coupling allows feedback of FRAPTRAN results in real time. Because of the significant impact of fuel models on key safety parameters such as PCT, detailed and accurate fuel models should be employed when evaluating PCT in LOCA analysis. It is noteworthy that the ECCS evaluation models laid out in the Appendix K to 10CFR50 require a provision for predicting cladding swelling and rupture and require to assume that the inside of the cladding react with steam after the rupture. The metal-water reaction energy can have significantly large effect on the reflood PCT, especially when fuel failure occurs. Effects of applying an advanced fuel model on the PCT evaluation can be clearly seen when comparing the MARS and the FRAPTRAN results in both the one-way calculation and the feedback calculation. As long as MARS and FRAPTRAN are used respectively in the ranges where they have been validated, the coupled calculation results are expected to be valid and to reveal various aspects of phenomena which have not been discovered in previous uncoupled calculations by MARS or FRAPTRAN.

  2. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  3. An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

    International Nuclear Information System (INIS)

    Hall, D.G.; Watkins, J.C.

    1987-01-01

    This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

  4. Intrauterine Intervention for the Treatment of Fetal Growth Restriction.

    Science.gov (United States)

    Spiroski, A-M; Oliver, M H; Harding, J E; Bloomfield, F H

    2016-01-01

    Fetal growth restriction (FGR) is associated with an increased incidence of fetal and neonatal death, and of neonatal morbidity. Babies born following FGR also are at risk of a range of postnatal complications, which may contribute to an increased incidence of disease later in life. There currently are no effective clinical interventions which improve perinatal survival, intrauterine growth and later outcomes of the FGR baby. Postnatal interventions aimed at promoting or accelerating growth in FGR babies to improve outcome, particularly neurodevelopmental outcomes, may further increase the risk of metabolic dysregulation and, therefore, the risk of developing chronic disease in adulthood. An intrauterine intervention to improve nutrition and growth in the FGR fetus may have the potential to decrease mortality and improve long-term outcomes by delaying preterm delivery and mitigating the need for and risks of accelerated postnatal growth.

  5. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  6. FFT-BM, Code Accuracy Evaluations with the 1D Fast Fourier Transform (FFT) Methodology

    International Nuclear Information System (INIS)

    D'Auria, F.

    2004-01-01

    1 - Description of program or function: FFT-BM is an integrated version of the programs package performing code accuracy evaluations with the 1D Fast Fourier Transform (FFT) methodology. It contains two programs: - CASEM: Takes care of the complete manipulation of data in order to evaluate the quantities through which the FFT method quantifies the code accuracy. - AAWFTO completes the evaluation of the average accuracy (AA) and related weighted frequency (WF) values in order to obtain the AAtot and WFtot values characterising the global calculation performance. 2 - Methods: The Fast Fourier Transform, or FFT, which is based on the Fourier analysis method is an optimised method for calculating the amplitude Vs frequency, of functions or experimental or computed data. In order to apply this methodology, after selecting the parameters to be analyzed, it is necessary to choose the following parameters: - number of curves (exp + calc) to be analyzed; - number of time windows to be analyzed; - sampling frequency; - cut frequency; - time begin and time end of each time window. 3 - Restrictions on the complexity of the problem: Up to 30 curves (exp + calc) and 5 time windows may be analyzed

  7. Evaluating the benefits of commercial building energy codes and improving federal incentives for code adoption.

    Science.gov (United States)

    Gilbraith, Nathaniel; Azevedo, Inês L; Jaramillo, Paulina

    2014-12-16

    The federal government has the goal of decreasing commercial building energy consumption and pollutant emissions by incentivizing the adoption of commercial building energy codes. Quantitative estimates of code benefits at the state level that can inform the size and allocation of these incentives are not available. We estimate the state-level climate, environmental, and health benefits (i.e., social benefits) and reductions in energy bills (private benefits) of a more stringent code (ASHRAE 90.1-2010) relative to a baseline code (ASHRAE 90.1-2007). We find that reductions in site energy use intensity range from 93 MJ/m(2) of new construction per year (California) to 270 MJ/m(2) of new construction per year (North Dakota). Total annual benefits from more stringent codes total $506 million for all states, where $372 million are from reductions in energy bills, and $134 million are from social benefits. These total benefits range from $0.6 million in Wyoming to $49 million in Texas. Private benefits range from $0.38 per square meter in Washington State to $1.06 per square meter in New Hampshire. Social benefits range from $0.2 per square meter annually in California to $2.5 per square meter in Ohio. Reductions in human/environmental damages and future climate damages account for nearly equal shares of social benefits.

  8. Training and support to improve ICD coding quality: A controlled before-and-after impact evaluation

    Directory of Open Access Journals (Sweden)

    Robin Dyers

    2017-06-01

    Full Text Available Background. The proposed National Health Insurance policy for South Africa (SA requires hospitals to maintain high-quality International Statistical Classification of Diseases (ICD codes for patient records. While considerable strides had been made to improve ICD coding coverage by digitising the discharge process in the Western Cape Province, further intervention was required to improve data quality. The aim of this controlled before-and-after study was to evaluate the impact of a clinician training and support initiative to improve ICD coding quality. Objective. To compare ICD coding quality between two central hospitals in the Western Cape before and after the implementation of a training and support initiative for clinicians at one of the sites. Methods. The difference in differences in data quality between the intervention site and the control site was calculated. Multiple logistic regression was also used to determine the odds of data quality improvement after the intervention and to adjust for potential differences between the groups. Results. The intervention had a positive impact of 38.0% on ICD coding completeness over and above changes that occurred at the control site. Relative to the baseline, patient records at the intervention site had a 6.6 (95% confidence interval 3.5 - 16.2 adjusted odds ratio of having a complete set of ICD codes for an admission episode after the introduction of the training and support package. The findings on impact on ICD coding accuracy were not significant. Conclusion. There is sufficient pragmatic evidence that a training and support package will have a considerable positive impact on ICD coding completeness in the SA setting.

  9. Application of dose evaluation of the MCNP code for interim spent fuel cask storage facility

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Iimoto, Takeshi; Ishikawa, Satoshi; Tsuboi, Takafumi; Teramura, Masahiro; Okamura, Tomomi; Narumiya, Yoshiyuki

    2007-01-01

    The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results. (author)

  10. Development and evaluation of a Naïve Bayesian model for coding causation of workers' compensation claims.

    Science.gov (United States)

    Bertke, S J; Meyers, A R; Wurzelbacher, S J; Bell, J; Lampl, M L; Robins, D

    2012-12-01

    Tracking and trending rates of injuries and illnesses classified as musculoskeletal disorders caused by ergonomic risk factors such as overexertion and repetitive motion (MSDs) and slips, trips, or falls (STFs) in different industry sectors is of high interest to many researchers. Unfortunately, identifying the cause of injuries and illnesses in large datasets such as workers' compensation systems often requires reading and coding the free form accident text narrative for potentially millions of records. To alleviate the need for manual coding, this paper describes and evaluates a computer auto-coding algorithm that demonstrated the ability to code millions of claims quickly and accurately by learning from a set of previously manually coded claims. The auto-coding program was able to code claims as a musculoskeletal disorders, STF or other with approximately 90% accuracy. The program developed and discussed in this paper provides an accurate and efficient method for identifying the causation of workers' compensation claims as a STF or MSD in a large database based on the unstructured text narrative and resulting injury diagnoses. The program coded thousands of claims in minutes. The method described in this paper can be used by researchers and practitioners to relieve the manual burden of reading and identifying the causation of claims as a STF or MSD. Furthermore, the method can be easily generalized to code/classify other unstructured text narratives. Published by Elsevier Ltd.

  11. 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.

  12. Evaluation of the methodology for dose calculation in microdosimetry with electrons sources using the MCNP5 Code

    International Nuclear Information System (INIS)

    Cintra, Felipe Belonsi de

    2010-01-01

    This study made a comparison between some of the major transport codes that employ the Monte Carlo stochastic approach in dosimetric calculations in nuclear medicine. We analyzed in detail the various physical and numerical models used by MCNP5 code in relation with codes like EGS and Penelope. The identification of its potential and limitations for solving microdosimetry problems were highlighted. The condensed history methodology used by MCNP resulted in lower values for energy deposition calculation. This showed a known feature of the condensed stories: its underestimates both the number of collisions along the trajectory of the electron and the number of secondary particles created. The use of transport codes like MCNP and Penelope for micrometer scales received special attention in this work. Class I and class II codes were studied and their main resources were exploited in order to transport electrons, which have particular importance in dosimetry. It is expected that the evaluation of available methodologies mentioned here contribute to a better understanding of the behavior of these codes, especially for this class of problems, common in microdosimetry. (author)

  13. The European source-term evaluation code ASTEC: status and applications, including CANDU plant applications

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Giordano, P.; Kissane, M.P.; Montanelli, T.; Schwinges, B.; Ganju, S.; Dickson, L.

    2004-01-01

    Research on light-water reactor severe accidents (SA) is still required in a limited number of areas in order to confirm accident-management plans. Thus, 49 European organizations have linked their SA research in a durable way through SARNET (Severe Accident Research and management NETwork), part of the European 6th Framework Programme. One goal of SARNET is to consolidate the integral code ASTEC (Accident Source Term Evaluation Code, developed by IRSN and GRS) as the European reference tool for safety studies; SARNET efforts include extending the application scope to reactor types other than PWR (including VVER) such as BWR and CANDU. ASTEC is used in IRSN's Probabilistic Safety Analysis level 2 of 900 MWe French PWRs. An earlier version of ASTEC's SOPHAEROS module, including improvements by AECL, is being validated as the Canadian Industry Standard Toolset code for FP-transport analysis in the CANDU Heat Transport System. Work with ASTEC has also been performed by Bhabha Atomic Research Centre, Mumbai, on IPHWR containment thermal hydraulics. (author)

  14. Evaluation of control room habitability in case of LOCA for Maanshan NPP using codes RADTRAD, HABIT and ALOHA

    International Nuclear Information System (INIS)

    Hsu, Wen-Sheng; Wang, Jong-Rong; Chen, Hsiung-Chih; Chiang, Yu; Chen, Shao-Wen; Shih, Chunkuan

    2018-01-01

    The method for the evaluation of the control room habitability is presented in this paper with focus on Maanshan PWR nuclear power plant (NPP) using the codes RADTRAD, HABIT, and ALOHA. Therefore, this paper is divided into two parts: The first part is the evaluation of the cumulative dose at the control room, the exclusion area boundary (EAB) and the low population zone (LPZ) in case of an design basis loss of coolant accident (DBA/LOCA). For this first part, the Maanshan NPP models of the code RADTRAD/SNAP were used for the analysis. The second part is the evaluation of the control room habitability under the assumption of CO 2 storage burst. For this part, the HABIT and ALOHA codes were used. As result it was seen that the RADTRAD calculation results are below the failure criteria of standard review plan (SRP) and 10 CFR 100.11. The HABIT and ALOHA results are below the R.G. 1.78 failure criteria. These results indicate that Maanshan NPP' habitability can be maintained under the above conditions.

  15. Evaluation of flaws in ferritic piping: ASME Code Appendix J, Deformation Plasticity Failure Assessment Diagram (DPFAD)

    International Nuclear Information System (INIS)

    Bloom, J.M.

    1991-08-01

    This report summarizes the methods and bases used by an ASME Code procedure for the evaluation of flaws in ferritic piping. The procedure is currently under consideration by the ASME Boiler and Pressure Vessel Code Committee of Section 11. The procedure was initially proposed in 1985 for the evaluation of the acceptability of flaws detected in piping during in-service inspection for certain materials, identified in Article IWB-3640 of the ASME Boiler and Pressure Vessel Code Section 11 ''Rules for In-service Inspection of Nuclear Power Plant Components.'' for which the fracture toughness is not sufficiently high to justify acceptance based solely on the plastic limit load evaluation methodology of Appendix C and IWB-3641. The procedure, referred to as Appendix J, originally included two approaches: a J-integral based tearing instability (J-T) analysis and the deformation plasticity failure assessment diagram (DPFAD) methodology. In Appendix J, a general DPFAD approach was simplified for application to part-through wall flows in ferritic piping through the use of a single DPFAD curve for circumferential flaws. Axial flaws are handled using two DPFAD curves where the ratio of flaw depth to wall thickness is used to determine the appropriate DPFAD curve. Flaws are evaluated in Appendix J by comparing the actual pipe applied stress with the allowable stress with the appropriate safety factors for the flaw size at the end of the evaluation period. Assessment points for circumferential and axial flaws are plotted on the appropriate failure assessment diagram. In addition, this report summarizes the experimental test predictions of the results of the Battelle Columbus Laboratory experiments, the Eiber experiments, and the JAERI tests using the Appendix J DPFAD methodology. Lastly, this report also provides guidelines for handling residual stresses in the evaluation procedure. 22 refs., 13 figs., 5 tabs

  16. Comparative gene expression profiling of placentas from patients with severe pre-eclampsia and unexplained fetal growth restriction

    Directory of Open Access Journals (Sweden)

    Kurahashi Hiroki

    2011-08-01

    Full Text Available Abstract Background It has been well documented that pre-eclampsia and unexplained fetal growth restriction (FGR have a common etiological background, but little is known about their linkage at the molecular level. The aim of this study was to further investigate the mechanisms underlying pre-eclampsia and unexplained FGR. Methods We analyzed differentially expressed genes in placental tissue from severe pre-eclamptic pregnancies (n = 8 and normotensive pregnancies with or (n = 8 without FGR (n = 8 using a microarray method. Results A subset of the FGR samples showed a high correlation coefficient overall in the microarray data from the pre-eclampsia samples. Many genes that are known to be up-regulated in pre-eclampsia are also up-regulated in FGR, including the anti-angiogenic factors, FLT1 and ENG, believed to be associated with the onset of maternal symptoms of pre-eclampsia. A total of 62 genes were found to be differentially expressed in both disorders. However, gene set enrichment analysis for these differentially expressed genes further revealed higher expression of TP53-downstream genes in pre-eclampsia compared with FGR. TP53-downstream apoptosis-related genes, such as BCL6 and BAX, were found to be significantly more up-regulated in pre-eclampsia than in FGR, although the caspases are expressed at equivalent levels. Conclusions Our current data indicate a common pathophysiology for FGR and pre-eclampsia, leading to an up-regulation of placental anti-angiogenic factors. However, our findings also suggest that it may possibly be the excretion of these factors into the maternal circulation through the TP53-mediated early-stage apoptosis of trophoblasts that leads to the maternal symptoms of pre-eclampsia.

  17. Evaluation of radiological dispersion/consequence codes supporting DOE nuclear facility SARs

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Paik, I.K.; Chung, D.Y.

    1996-01-01

    Since the early 1990s, the authorization basis documentation of many U.S. Department of Energy (DOE) nuclear facilities has been upgraded to comply with DOE orders and standards. In this process, many safety analyses have been revised. Unfortunately, there has been nonuniform application of software, and the most appropriate computer and engineering methodologies often are not applied. A DOE Accident Phenomenology and Consequence (APAC) Methodology Evaluation Program was originated at the request of DOE Defense Programs to evaluate the safety analysis methodologies used in nuclear facility authorization basis documentation and to define future cost-effective support and development initiatives. Six areas, including source term development (fire, spills, and explosion analysis), in-facility transport, and dispersion/ consequence analysis (chemical and radiological) are contained in the APAC program. The evaluation process, codes considered, key results, and recommendations for future model and software development of the Radiological Dispersion/Consequence Working Group are summarized in this paper

  18. Genetic markers for inherited thrombophilia are associated with fetal growth retardation in the population of Central Russia.

    Science.gov (United States)

    Reshetnikov, Evgeny; Zarudskaya, Oksana; Polonikov, Alexey; Bushueva, Olga; Orlova, Valentina; Krikun, Evgeny; Dvornyk, Volodymyr; Churnosov, Mikhail

    2017-07-01

    The aim of this study was to examine the role of hereditary thrombophilia in the development of fetal growth retardation (FGR) in the population of Central Russia. The case-control study sample included 497 women in the third trimester of pregnancy recruited during 2009-2013. The participants were enrolled into two groups: patients with FGR (n = 250) and controls without FGR (n = 247). The participants were genotyped for four genetic markers of hereditary thrombophilia: factor V Leiden (G > A FV, rs6025), prothrombin (G > A FII, rs1799963), factor VII (G > A FVII, rs6046), and fibrinogen (G > A FI, rs1800790). The genetic factors for an increased risk of FGR were allele G of rs6046 (odds ratio [OR] = 2.34) and genotype GG of rs6046 (OR = 2.64), whereas genotype GA of rs6046 had the protective value (OR = 0.42). A combination of alleles G of rs1799963, A of rs6046, and G of rs1800790 (OR = 0.31) reduces the risk of FGR. Polymorphism rs6046 of the FVII gene is associated with the development of FGR. © 2017 Japan Society of Obstetrics and Gynecology.

  19. Evaluation code for the dose due to the discharges of liquid effluents of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Lopez, Fabio O.; Boutet, Luis I.; Bruno, Hector A.; Gavini, Ricardo M.

    2004-01-01

    A new methodology is presented to assess the evaluation of the radiological impact to the population, due to the discharges to the environment of liquids effluents of Central Nuclear Embalse (CNE), located in the Province of Cordoba (Argentina). In order to carry out the dose evaluation, a code denominated EDDELIQ was developed, in this code the calculation of the radionuclides concentration in the water lake is made by means of a simple physical model of the type of complete mixture. The physical model is solved numerically by means of Runge Kutta method of second order. (author)

  20. Performance Evaluation of Wavelet-Coded OFDM on a 4.9 Gbps W-Band Radio-over-Fiber Link

    DEFF Research Database (Denmark)

    Cavalcante, Lucas Costa Pereira; Rommel, Simon; Dinis, Rui

    2017-01-01

    Future generation mobile communications running on mm-wave frequencies will require great robustness against frequency selective channels. In this work we evaluate the transmission performance of 4.9 Gbps Wavelet-Coded OFDM signals on a 10 km fiber plus 58 m wireless Radio-over-Fiber link using...... a mm-wave radio frequency carrier. The results show that a 2×128 Wavelet-Coded OFDM system achieves a bit-error rate of 1e-4 with nearly 2.5 dB less signal-to-noise ratio than a convolutional coded OFDM system with equivalent spectral efficiency for 8 GHz-wide signals with 512 sub-carriers on a carrier...

  1. Development of ECOREA-II code for the evaluation of exposures from radionuclides through food Chain

    International Nuclear Information System (INIS)

    Yu, Dong Han; Lee, Han Soo

    2002-01-01

    The release of radionuclides from nuclear facilities following an accident into air results in human exposures by intakes of plant products such as rice, vegetables and/or animal products including meat, milk and eggs from contaminated soil. In order to evaluate such exposures from radioactive substances, it is essential to mathematically predict the behavior of these substances in the environments. A computer code, named 'ECOREA-II' is developing to assess human exposures through food chain of such substances in Korea. ECOREA-II code has a dynamic compartment-based model at its core, the graphical user interface (GUI) for the selection of input parameters and result displays on personal computers, and generation of data files for a GIS (Graphical Information System). Even the code is developed mostly based currently available models and/or codes, a new model is included for the time-dependent growth dilution in a vegetation part. Effort on The development of the code is towards the prediction of the behavior and pattern of radionuclides in a specific food chain condition in Korea. Finally, it provides a more user-friendly environment such as GUI developed based on the VBA(Visual Basic Application) for personal users. Therefore, the current code, when more fully developed, is expected to increase the understanding of environmental safety assessment of nuclear facilities following an accident and provide a reasonable regulatory guideline with respect to food safety issues

  2. Prediction of fetal growth restriction using estimated fetal weight vs a combined screening model in the third trimester.

    Science.gov (United States)

    Miranda, J; Rodriguez-Lopez, M; Triunfo, S; Sairanen, M; Kouru, H; Parra-Saavedra, M; Crovetto, F; Figueras, F; Crispi, F; Gratacós, E

    2017-11-01

    To compare the performance of third-trimester screening, based on estimated fetal weight centile (EFWc) vs a combined model including maternal baseline characteristics, fetoplacental ultrasound and maternal biochemical markers, for the prediction of small-for-gestational-age (SGA) neonates and late-onset fetal growth restriction (FGR). This was a nested case-control study within a prospective cohort of 1590 singleton gestations undergoing third-trimester (32 + 0 to 36 + 6 weeks' gestation) evaluation. Maternal baseline characteristics, mean arterial pressure, fetoplacental ultrasound and circulating biochemical markers (placental growth factor (PlGF), lipocalin-2, unconjugated estriol and inhibin A) were assessed in all women who subsequently delivered a SGA neonate (n = 175), defined as birth weight < 10 th centile according to customized standards, and in a control group (n = 875). Among SGA cases, those with birth weight < 3 rd centile and/or abnormal uterine artery pulsatility index (UtA-PI) and/or abnormal cerebroplacental ratio (CPR) were classified as FGR. Logistic regression predictive models were developed for SGA and FGR, and their performance was compared with that obtained using EFWc alone. In SGA cases, EFWc, CPR Z-score and maternal serum concentrations of unconjugated estriol and PlGF were significantly lower, while mean UtA-PI Z-score and lipocalin-2 and inhibin A concentrations were significantly higher, compared with controls. Using EFWc alone, 52% (area under receiver-operating characteristics curve (AUC), 0.82 (95% CI, 0.77-0.85)) of SGA and 64% (AUC, 0.86 (95% CI, 0.81-0.91)) of FGR cases were predicted at a 10% false-positive rate. A combined screening model including a-priori risk (maternal characteristics), EFWc, UtA-PI, PlGF and estriol (with lipocalin-2 for SGA) achieved a detection rate of 61% (AUC, 0.86 (95% CI, 0.83-0.89)) for SGA cases and 77% (AUC, 0.92 (95% CI, 0.88-0.95)) for FGR. The combined model for the

  3. The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL

    International Nuclear Information System (INIS)

    Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.

    1983-03-01

    This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)

  4. Safety evaluation of liquid radioactive effluents treatment system in a BWR reactor, through the LIQM03 code

    International Nuclear Information System (INIS)

    Zorrilla R, S.H.

    1978-01-01

    In this work we made a safety evaluation of the liquid radioactive effluents system in a plant using a BWR similar to that now installed in Laguna Verde. For that purpose, the computation program ORIGENwas modified, in order to keep up to date and adapt it to the PDP 10 computer, which is operating at the Computation Department of the Nuclear Center of Mexico, the code LIQM03 was the result of this modification. As usual in this work we dealt with problems which were solved opportunely, now we have at our disposal the code LIQM03 which will be in the future a very useful tool for this kind of evaluations. (author)

  5. ANCON: A code for the evaluation of complex fault trees in personal computers

    International Nuclear Information System (INIS)

    Napoles, J.G.; Salomon, J.; Rivero, J.

    1990-01-01

    Performing probabilistic safety analysis has been recognized worldwide as one of the more effective ways for further enhancing safety of Nuclear Power Plants. The evaluation of fault trees plays a fundamental role in these analysis. Some existing limitations in RAM and execution speed of personal computers (PC) has restricted so far their use in the analysis of complex fault trees. Starting from new approaches in the data structure and other possibilities the ANCON code can evaluate complex fault trees in a PC, allowing the user to do a more comprehensive analysis of the considered system in reduced computing time

  6. Evaluation of accuracy of Monte Carlo code MVP with VHTRC experiments. Multiplication factor at criticality, burnable poison worth and void worth

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Yamashita, Kiyonobu; Fiujimoto, Nozomu; Nakano, Masaaki , Yamane, Tsuyoshi; Akino, Fujiyoshi.

    1997-11-01

    Experimental data of VHTRC (Very High Temperature Reactor Critical Assembly) were analyzed using Monte Carlo code MVP (general purpose Monte Carlo code of neutron and photon transport calculations based on the continuous energy method). The calculation accuracy of the code was evaluated by the analysis for nuclear characteristics of a HTGR (high temperature gas-cooled reactor). The MVP code can analyze with a detailed three-dimensional core model with a few approximations. The HTGRs have following characteristics from view point of nuclear design : they have burnable poisons, many void holes, namely, the control insertion holes and so on. Taking account of these characteristics, multiplication factor at criticality, burnable poison worth, and void worth were evaluated. The maximum calculation errors were 0.8%Δk, 7%, and 25% respectively, From these results, it can be concluded that the MVP code is able to be applied to the nuclear characteristics analysis of the HTGR like the High Temperature Engineering Test Reactor (HTTR). (author)

  7. PL-MOD: a computer code for modular fault tree analysis and evaluation

    International Nuclear Information System (INIS)

    Olmos, J.; Wolf, L.

    1978-01-01

    The computer code PL-MOD has been developed to implement the modular methodology to fault tree analysis. In the modular approach, fault tree structures are characterized by recursively relating the top tree event to all basic event inputs through a set of equations, each defining an independent modular event for the tree. The advantages of tree modularization lie in that it is a more compact representation than the minimal cut-set description and in that it is well suited for fault tree quantification because of its recursive form. In its present version, PL-MOD modularizes fault trees and evaluates top and intermediate event failure probabilities, as well as basic component and modular event importance measures, in a very efficient way. Thus, its execution time for the modularization and quantification of a PWR High Pressure Injection System reduced fault tree was 25 times faster than that necessary to generate its equivalent minimal cut-set description using the computer code MOCUS

  8. Coding for urologic office procedures.

    Science.gov (United States)

    Dowling, Robert A; Painter, Mark

    2013-11-01

    This article summarizes current best practices for documenting, coding, and billing common office-based urologic procedures. Topics covered include general principles, basic and advanced urologic coding, creation of medical records that support compliant coding practices, bundled codes and unbundling, global periods, modifiers for procedure codes, when to bill for evaluation and management services during the same visit, coding for supplies, and laboratory and radiology procedures pertinent to urology practice. Detailed information is included for the most common urology office procedures, and suggested resources and references are provided. This information is of value to physicians, office managers, and their coding staff. Copyright © 2013 Elsevier Inc. All rights reserved.

  9. Tc-99m DTPA perfusion scintigraphy and color coded duplex sonography in the evaluation of minimal renal allograft perfusion

    International Nuclear Information System (INIS)

    Bair, H.J.; Platsch, G.; Wolf, F.; Guenter, E.; Becker, D.; Rupprecht, H.; Neumayer, H.H.

    1997-01-01

    Aim: The clinical impact of perfusion scintigraphy versus color coded Duplex sonography was evaluated, with respect to their potential in assessing minimal allograft perfusion in vitally threatened kidney transplants, i.e. oligoanuric allografts suspected to have either severe rejection or thrombosis of the renal vein or artery. Methods: From July 1990 to August 1994 the grafts of 15 out of a total of 315 patients were vitally threatened. Technetium-99m DTPA scintigraphy and color coded Duplex sonography were performed in all patients. For scintigraphic evaluation of transplant perfusion analog scans up to 60 min postinjection, and time-activity curves over the first 60 sec after injection of 370-440 MBq Tc-99m diethylenetriaminepentaacetate acid (DTPA) were used and classified by a perfusion score, the time between renal and iliac artery peaks (TDiff) and the washout of the renogram curve. Additionally, evaluation of excretion function and assessment of vascular or urinary leaks were performed. By color coded Duplex sonography the perfusion in all sections of the graft as well as the vascular anastomoses were examined and the maximal blood flow velocity (Vmax) and the resistive index (RI) in the renal artery were determined by means of the pulsed Doppler device. Pathologic-anatomical diagnosis was achieved by either biopsy or post-explant histology in all grafts. Results: Scintigraphy and color coded Duplex sonography could reliably differentiate minimal (8/15) and not perfused (7/15) renal allografts. The results were confirmed either by angiography in digital subtraction technique (DSA) or the clinical follow up. Conclusion: In summary, perfusion scintigraphy and color coded Duplex sonography are comparable modalities to assess kidney graft perfusion. In clinical practice scintigraphy and colorcoded Doppler sonography can replace digital subtraction angiography in the evaluation of minimal allograft perfusion. (orig.) [de

  10. Random linear codes in steganography

    Directory of Open Access Journals (Sweden)

    Kamil Kaczyński

    2016-12-01

    Full Text Available Syndrome coding using linear codes is a technique that allows improvement in the steganographic algorithms parameters. The use of random linear codes gives a great flexibility in choosing the parameters of the linear code. In parallel, it offers easy generation of parity check matrix. In this paper, the modification of LSB algorithm is presented. A random linear code [8, 2] was used as a base for algorithm modification. The implementation of the proposed algorithm, along with practical evaluation of algorithms’ parameters based on the test images was made.[b]Keywords:[/b] steganography, random linear codes, RLC, LSB

  11. Performance Evaluation of Spectral Amplitude Codes for OCDMA PON

    DEFF Research Database (Denmark)

    Binti Othman, Maisara; Jensen, Jesper Bevensee; Zhang, Xu

    2011-01-01

    the MAI effects in OCDMA. The performance has been characterized through received optical power (ROP) sensitivity and dispersion tolerance assessments. The numerical results show that the ZCC code has a slightly better performance compared to the other two codes for the ROP and similar behavior against...

  12. Levels of serum-circulating angiogenic factors within 1 week prior to delivery are closely related to conditions of pregnant women with pre-eclampsia, gestational hypertension, and/or fetal growth restriction.

    Science.gov (United States)

    Nanjo, Sakiko; Minami, Sawako; Mizoguchi, Mika; Yamamoto, Madoka; Yahata, Tamaki; Toujima, Saori; Shiro, Michihisa; Kobayashi, Aya; Muragaki, Yasuteru; Ino, Kazuhiko

    2017-12-01

    We aimed to investigate maternal serum angiogenic marker profiles within 1 week prior to delivery in cases of gestational hypertension (GH), pre-eclampsia (PE), and/or fetal growth restriction (FGR) with different clinical conditions. We enrolled 165 women with singleton pregnancy. The participants were classified based on three characteristics: (i) proteinuria (GH and PE); (ii) FGR (PE with FGR [PE + FGR], PE alone, and FGR alone); and (iii) onset (early onset PE [EO PE] and late-onset PE [LO PE]). All sera were obtained within 1 week prior to delivery, and soluble fms-like tyrosine kinase 1 (sFlt-1), soluble endoglin (sEng), and placental growth factor (PlGF) were measured with enzyme-linked immunosorbent assay. (i) In PE, a significantly increased sFlt-1, sEng, and sFlt-1 to PlGF ratio (sFlt-1/PlGF) and significantly decreased PlGF were observed compared with GH and Term control, whereas in GH, only sFlt-1/PlGF was significantly higher than Term control. (ii) In PE + FGR, similar changes were more markedly shown compared with PE alone. The FGR alone group exhibited similar tendencies as PE, although significant differences were found in PlGF and sEng levels. (iii) In EO PE, significant changes were observed in all factors compared with LO PE or Term control, while no significant change in PlGF levels was observed between LO PE and Term control. We demonstrated that the levels of circulating angiogenic factors just before delivery are correlated with the severity of hypertensive disorders of pregnancy and FGR. Profiling these specific markers may contribute to better understanding of the clinical conditions in individual patients and their pathogenesis. © 2017 Japan Society of Obstetrics and Gynecology.

  13. RELAP5/MOD2 code assessment

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

  14. RELAP5/MOD2 code assessment

    Energy Technology Data Exchange (ETDEWEB)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Miller, F.R.

    1985-11-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

  15. GESDATA: A failure-data management code

    International Nuclear Information System (INIS)

    Garcia Gay, J.; Francia Gonzalez, L.; Ortega Prieto, P.; Mira McWilliams, J.; Aguinaga Zapata, M.

    1987-01-01

    GESDATA is a failure data management code for both qualitative and quantitative fault-tree evaluation. Data management using the code should provide the analyst, in the quickest and easiest way, with the reliability data which constitute the input values for fault-tree evaluation programs. (orig./HSCH)

  16. Development of safety evaluation methods and analysis codes applied to the safety regulations for the design and construction stage of fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purposes of this study are to develop the safety evaluation methods and analysis codes needed in the design and construction stage of fast breeder reactor (FBR). In JFY 2012, the following results are obtained. As for the development of safety evaluation methods needed in the safety examination conducted for the reactor establishment permission, development of the analysis codes, such as core damage analysis code, were carried out following the planned schedule. As for the development of the safety evaluation method needed for the risk informed safety regulation, the quantification technique of the event tree using the Continuous Markov chain Monte Carlo method (CMMC method) were studied. (author)

  17. Trophoblastic progranulin expression is upregulated in cases of fetal growth restriction and preeclampsia.

    Science.gov (United States)

    Stubert, Johannes; Schattenberg, Florian; Richter, Dagmar-Ulrike; Dieterich, Max; Briese, Volker

    2012-05-13

    The expression of the anti-inflammatory glycoprotein progranulin and the hypoxia-induced transcription factor 1α (HIF-1α) in the villous trophoblast was compared between placentae from patients with preeclampsia (PE), fetal growth restriction (FGR), and normal controls. Matched pairs analysis of third trimester placentae specimens (mean gestational age 36+2) was performed by semiquantitative measurements of the immunohistochemical staining intensities for progranulin and HIF-1α expression (PE n=13, FGR n=9 and controls n=11). Further, placental progranulin mRNA expression was analyzed by qRT-PCR on term placentae (n=3 for each group). Compared to controls, villous trophoblast revealed a significantly higher expression of progranulin in cases of PE (Pprogranulin protein was not accompanied by an increase of the progranulin mRNA in term placentae. Increased expression of progranulin protein in villous trophoblast cells in cases of PE and FGR may result from disturbed placental development and, therefore, may be of pathogenetic importance. The increase was correlated to HIF-1α expression. Further evaluation of this potential mechanism of regulation is required.

  18. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  19. Empirical Evaluation of Superposition Coded Multicasting for Scalable Video

    KAUST Repository

    Chun Pong Lau

    2013-03-01

    In this paper we investigate cross-layer superposition coded multicast (SCM). Previous studies have proven its effectiveness in exploiting better channel capacity and service granularities via both analytical and simulation approaches. However, it has never been practically implemented using a commercial 4G system. This paper demonstrates our prototype in achieving the SCM using a standard 802.16 based testbed for scalable video transmissions. In particular, to implement the superposition coded (SPC) modulation, we take advantage a novel software approach, namely logical SPC (L-SPC), which aims to mimic the physical layer superposition coded modulation. The emulation results show improved throughput comparing with generic multicast method.

  20. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Miscellaneous -- Volume 3, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, L.M.; Jordon, W.C. [Oak Ridge National Lab., TN (United States); Edwards, A.L. [Oak Ridge National Lab., TN (United States)]|[Lawrence Livermore National Lab., CA (United States)] [and others

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice; (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System developments has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. This manual is divided into three volumes: Volume 1--for the control module documentation, Volume 2--for the functional module documentation, and Volume 3--for the data libraries and subroutine libraries.

  1. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Miscellaneous -- Volume 3, Revision 4

    International Nuclear Information System (INIS)

    Petrie, L.M.; Jordon, W.C.; Edwards, A.L.

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice; (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System developments has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. This manual is divided into three volumes: Volume 1--for the control module documentation, Volume 2--for the functional module documentation, and Volume 3--for the data libraries and subroutine libraries

  2. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  3. Building Energy Efficiency in India: Compliance Evaluation of Energy Conservation Building Code

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sha; Evans, Meredydd; Delgado, Alison

    2014-03-26

    s enactment, only two states and one territory out of 35 Indian states and union territories formally adopted ECBC and six additional states are in the legislative process of approving ECBC. There are several barriers that slow down the process. First, stakeholders, such as architects, developers, and state and local governments, lack awareness of building energy efficiency, and do not have enough capacity and resources to implement ECBC. Second, institution for implementing ECBC is not set up yet; ECBC is not included in local building by-laws or incorporated into the building permit process. Third, there is not a systematic approach to measuring and verifying compliance and energy savings, and thus the market does not have enough confidence in ECBC. Energy codes achieve energy savings only when projects comply with codes, yet only few countries measure compliance consistently and periodic checks often indicate poor compliance in many jurisdictions. China and the U.S. appear to be two countries with comprehensive systems in code enforcement and compliance The United States recently developed methodologies measuring compliance with building energy codes at the state level. China has an annual survey investigating code compliance rate at the design and construction stages in major cities. Like many developing countries, India has only recently begun implementing an energy code and would benefit from international experience on code compliance. In this paper, we examine lessons learned from the U.S. and China on compliance assessment and how India can apply these lessons to develop its own compliance evaluation approach. This paper also provides policy suggestions to national, state, and local governments to improve compliance and speed up ECBC implementation.

  4. Impacts of Model Building Energy Codes

    Energy Technology Data Exchange (ETDEWEB)

    Athalye, Rahul A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sivaraman, Deepak [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bartlett, Rosemarie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-10-31

    The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) periodically evaluates national and state-level impacts associated with energy codes in residential and commercial buildings. Pacific Northwest National Laboratory (PNNL), funded by DOE, conducted an assessment of the prospective impacts of national model building energy codes from 2010 through 2040. A previous PNNL study evaluated the impact of the Building Energy Codes Program; this study looked more broadly at overall code impacts. This report describes the methodology used for the assessment and presents the impacts in terms of energy savings, consumer cost savings, and reduced CO2 emissions at the state level and at aggregated levels. This analysis does not represent all potential savings from energy codes in the U.S. because it excludes several states which have codes which are fundamentally different from the national model energy codes or which do not have state-wide codes. Energy codes follow a three-phase cycle that starts with the development of a new model code, proceeds with the adoption of the new code by states and local jurisdictions, and finishes when buildings comply with the code. The development of new model code editions creates the potential for increased energy savings. After a new model code is adopted, potential savings are realized in the field when new buildings (or additions and alterations) are constructed to comply with the new code. Delayed adoption of a model code and incomplete compliance with the code’s requirements erode potential savings. The contributions of all three phases are crucial to the overall impact of codes, and are considered in this assessment.

  5. Simulation of integral local tests with high-burnup fuel

    International Nuclear Information System (INIS)

    Gyori, G.

    2011-01-01

    The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)

  6. Twenty years of fracture mechanics and flaw evaluation applications in the ASME Nuclear Code

    International Nuclear Information System (INIS)

    Riccardella, P.C.

    1991-01-01

    The paper presents a retrospective on the development and applications of fracture mechanics-based toughness requirements and flaw evaluation methodology in Sections III and XI of the ASME Code. Section III developments range from the rules and requirements for thick section Class 1 pressure vessels to thinner section components in other Classes. Section XI applications include flaw acceptance standards and evaluation methodology for various components ranging from pressure vessels to thins section piping of carbon and austenitic steels. The experience gained in operating plant applications of these rules and procedures are also discussed

  7. Survey of nuclear fuel-cycle codes

    International Nuclear Information System (INIS)

    Thomas, C.R.; de Saussure, G.; Marable, J.H.

    1981-04-01

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code

  8. Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

    International Nuclear Information System (INIS)

    Dozaki, Koji

    2007-01-01

    Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

  9. Altered decorin leads to disrupted endothelial cell function: a possible mechanism in the pathogenesis of fetal growth restriction?

    Science.gov (United States)

    Chui, A; Murthi, P; Gunatillake, T; Brennecke, S P; Ignjatovic, V; Monagle, P T; Whitelock, J M; Said, J M

    2014-08-01

    Fetal growth restriction (FGR) is a key cause of adverse pregnancy outcome where maternal and fetal factors are identified as contributing to this condition. Idiopathic FGR is associated with altered vascular endothelial cell functions. Decorin (DCN) has important roles in the regulation of endothelial cell functions in vascular environments. DCN expression is reduced in FGR. The objectives were to determine the functional consequences of reduced DCN in a human microvascular endothelial cell line model (HMVEC), and to determine downstream targets of DCN and their expression in primary placental microvascular endothelial cells (PLECs) from control and FGR-affected placentae. Short-interference RNA was used to reduce DCN expression in HMVECs and the effect on proliferation, angiogenesis and thrombin generation was determined. A Growth Factor PCR Array was used to identify downstream targets of DCN. The expression of target genes in control and FGR PLECs was performed. DCN reduction decreased proliferation and angiogenesis but increased thrombin generation with no effect on apoptosis. The array identified three targets of DCN: FGF17, IL18 and MSTN. Validation of target genes confirmed decreased expression of VEGFA, MMP9, EGFR1, IGFR1 and PLGF in HMVECs and PLECs from control and FGR pregnancies. Reduction of DCN in vascular endothelial cells leads to disrupted cell functions. The targets of DCN include genes that play important roles in angiogenesis and cellular growth. Therefore, differential expression of these may contribute to the pathogenesis of FGR and disease states in other microvascular circulations. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. Computer codes in particle transport physics

    International Nuclear Information System (INIS)

    Pesic, M.

    2004-01-01

    Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option

  11. Performance Evaluation of HARQ Technique with UMTS Turbo Code

    Directory of Open Access Journals (Sweden)

    S. S. Brkić

    2011-11-01

    Full Text Available The hybrid automatic repeat request technique (HARQ represents the error control principle which combines an error correcting code and automatic repeat request procedure (ARQ, within the same transmission system. In this paper, using Monte Carlo simulation process, the characteristics of HARQ technique are determined, for the case of the Universal Mobile Telecommunication System (UMTS turbo code.

  12. An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Allison, E-mail: lewis.allison10@gmail.com [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Smith, Ralph [Department of Mathematics, North Carolina State University, Raleigh, NC 27695 (United States); Williams, Brian [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Figueroa, Victor [Sandia National Laboratories, Albuquerque, NM 87185 (United States)

    2016-11-01

    For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is to employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.

  13. Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1980-01-01

    A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

  14. On the Evaluation of Pebble Bead Reactor Critical Experiments Using the Pebbed Code

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Sen, R. Sonat

    2014-01-01

    Critical experiments pose a particular but necessary challenge to validating pebble bed reactor design codes. Fuel and core heterogeneities, impurities in graphite, variable packing of pebbles, and moderately strong neutronic coupling are among the factors that inject uncertainty into the results obtained with lower fidelity core physics models. Some of these are addressed in this study. The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling

  15. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  16. PEAR code review

    International Nuclear Information System (INIS)

    De Wit, R.; Jamieson, T.; Lord, M.; Lafortune, J.F.

    1997-07-01

    As a necessary component in the continuous improvement and refinement of methodologies employed in the nuclear industry, regulatory agencies need to periodically evaluate these processes to improve confidence in results and ensure appropriate levels of safety are being achieved. The independent and objective review of industry-standard computer codes forms an essential part of this program. To this end, this work undertakes an in-depth review of the computer code PEAR (Public Exposures from Accidental Releases), developed by Atomic Energy of Canada Limited (AECL) to assess accidental releases from CANDU reactors. PEAR is based largely on the models contained in the Canadian Standards Association (CSA) N288.2-M91. This report presents the results of a detailed technical review of the PEAR code to identify any variations from the CSA standard and other supporting documentation, verify the source code, assess the quality of numerical models and results, and identify general strengths and weaknesses of the code. The version of the code employed in this review is the one which AECL intends to use for CANDU 9 safety analyses. (author)

  17. The 1989 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.

    1989-12-01

    This document summarizes the 1989 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  18. Evaluating Coding Accuracy in General Surgery Residents' Accreditation Council for Graduate Medical Education Procedural Case Logs.

    Science.gov (United States)

    Balla, Fadi; Garwe, Tabitha; Motghare, Prasenjeet; Stamile, Tessa; Kim, Jennifer; Mahnken, Heidi; Lees, Jason

    .0043). The survey response rate was 100%. Survey results indicated that inability to find the precise code within the ACGME search interface and unfamiliarity with available CPT codes were by far the most common perceived barriers to accuracy. Survey results also indicated that most residents (74%) believe that they code accurately most of the time and agree that their case log would accurately represent their operative experience (66.6%). This is the first study to evaluate correctness of residents' ACGME case logs in general surgery. The degree of inaccuracy found here necessitates further investigation into the etiology of these discrepancies. Instruction on coding practices should also benefit the residents after graduation. Optimizing communication among attendings and residents, improving ACGME coding search interface, and implementing consistent coding practices could improve accuracy giving a more realistic view of residents' operative experience. Published by Elsevier Inc.

  19. Implementation and evaluation of a simulation curriculum for paediatric residency programs including just-in-time in situ mock codes.

    Science.gov (United States)

    Sam, Jonathan; Pierse, Michael; Al-Qahtani, Abdullah; Cheng, Adam

    2012-02-01

    To develop, implement and evaluate a simulation-based acute care curriculum in a paediatric residency program using an integrated and longitudinal approach. Curriculum framework consisting of three modular, year-specific courses and longitudinal just-in-time, in situ mock codes. Paediatric residency program at BC Children's Hospital, Vancouver, British Columbia. The three year-specific courses focused on the critical first 5 min, complex medical management and crisis resource management, respectively. The just-in-time in situ mock codes simulated the acute deterioration of an existing ward patient, prepared the actual multidisciplinary code team, and primed the surrounding crisis support systems. Each curriculum component was evaluated with surveys using a five-point Likert scale. A total of 40 resident surveys were completed after each of the modular courses, and an additional 28 surveys were completed for the overall simulation curriculum. The highest Likert scores were for hands-on skill stations, immersive simulation environment and crisis resource management teaching. Survey results also suggested that just-in-time mock codes were realistic, reinforced learning, and prepared ward teams for patient deterioration. A simulation-based acute care curriculum was successfully integrated into a paediatric residency program. It provides a model for integrating simulation-based learning into other training programs, as well as a model for any hospital that wishes to improve paediatric resuscitation outcomes using just-in-time in situ mock codes.

  20. Evaluating training of screening, brief intervention, and referral to treatment (SBIRT) for substance use: Reliability of the MD3 SBIRT Coding Scale.

    Science.gov (United States)

    DiClemente, Carlo C; Crouch, Taylor Berens; Norwood, Amber E Q; Delahanty, Janine; Welsh, Christopher

    2015-03-01

    Screening, brief intervention, and referral to treatment (SBIRT) has become an empirically supported and widely implemented approach in primary and specialty care for addressing substance misuse. Accordingly, training of providers in SBIRT has increased exponentially in recent years. However, the quality and fidelity of training programs and subsequent interventions are largely unknown because of the lack of SBIRT-specific evaluation tools. The purpose of this study was to create a coding scale to assess quality and fidelity of SBIRT interactions addressing alcohol, tobacco, illicit drugs, and prescription medication misuse. The scale was developed to evaluate performance in an SBIRT residency training program. Scale development was based on training protocol and competencies with consultation from Motivational Interviewing coding experts. Trained medical residents practiced SBIRT with standardized patients during 10- to 15-min videotaped interactions. This study included 25 tapes from the Family Medicine program coded by 3 unique coder pairs with varying levels of coding experience. Interrater reliability was assessed for overall scale components and individual items via intraclass correlation coefficients. Coder pair-specific reliability was also assessed. Interrater reliability was excellent overall for the scale components (>.85) and nearly all items. Reliability was higher for more experienced coders, though still adequate for the trained coder pair. Descriptive data demonstrated a broad range of adherence and skills. Subscale correlations supported concurrent and discriminant validity. Data provide evidence that the MD3 SBIRT Coding Scale is a psychometrically reliable coding system for evaluating SBIRT interactions and can be used to evaluate implementation skills for fidelity, training, assessment, and research. Recommendations for refinement and further testing of the measure are discussed. (PsycINFO Database Record (c) 2015 APA, all rights reserved).

  1. Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

    International Nuclear Information System (INIS)

    Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

    2004-01-01

    Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

  2. Sudan-decoding generalized geometric Goppa codes

    DEFF Research Database (Denmark)

    Heydtmann, Agnes Eileen

    2003-01-01

    Generalized geometric Goppa codes are vector spaces of n-tuples with entries from different extension fields of a ground field. They are derived from evaluating functions similar to conventional geometric Goppa codes, but allowing evaluation in places of arbitrary degree. A decoding scheme...... for these codes based on Sudan's improved algorithm is presented and its error-correcting capacity is analyzed. For the implementation of the algorithm it is necessary that the so-called increasing zero bases of certain spaces of functions are available. A method to obtain such bases is developed....

  3. Supporting qualified database for V and V and uncertainty evaluation of best-estimate system codes

    International Nuclear Information System (INIS)

    Petruzzi, A.; D'Auria, F.

    2014-01-01

    Uncertainty evaluation constitutes a key feature of BEPU (Best Estimate Plus Uncertainty) process. The uncertainty can be the result of a Monte Carlo type analysis involving input uncertainty parameters or the outcome of a process involving the use of experimental data and connected code calculations. Those uncertainty methods are discussed in several papers and guidelines (IAEA-SRS- 52, OECD/NEA BEMUSE reports). The present paper aims at discussing the role and the depth of the analysis required for merging from one side suitable experimental data and on the other side qualified code calculation results. This aspect is mostly connected with the second approach for uncertainty mentioned above, but it can be used also in the framework of the first approach. Namely, the paper discusses the features and structure of the database that includes the following kinds of documents: 1. The' RDS-facility' (Reference Data Set for the selected facility): this includes the description of the facility, the geometrical characterization of any component of the facility, the instrumentations, the data acquisition system, the evaluation of pressure losses, the physical properties of the material and the characterization of pumps, valves and heat losses; 2. The 'RDS-test' (Reference Data Set for the selected test of the facility): this includes the description of the main phenomena investigated during the test, the configuration of the facility for the selected test (possible new evaluation of pressure and heat losses if needed) and the specific boundary and initial conditions; 3. The 'QP' (Qualification Report) of the code calculation results: this includes the description of the nodalization developed following a set of homogeneous techniques, the achievement of the steady state conditions and the qualitative and quantitative analysis of the transient with the characterization of the Relevant Thermal-Hydraulics Aspects (RTA); 4. The EH (Engineering

  4. An empirical evaluation of the US Beer Institute's self-regulation code governing the content of beer advertising.

    Science.gov (United States)

    Babor, Thomas F; Xuan, Ziming; Damon, Donna; Noel, Jonathan

    2013-10-01

    We evaluated advertising code violations using the US Beer Institute guidelines for responsible advertising. We applied the Delphi rating technique to all beer ads (n = 289) broadcast in national markets between 1999 and 2008 during the National Collegiate Athletic Association basketball tournament games. Fifteen public health professionals completed ratings using quantitative scales measuring the content of alcohol advertisements (e.g., perceived actor age, portrayal of excessive drinking) according to 1997 and 2006 versions of the Beer Institute Code. Depending on the code version, exclusion criteria, and scoring method, expert raters found that between 35% and 74% of the ads had code violations. There were significant differences among producers in the frequency with which ads with violations were broadcast, but not in the proportions of unique ads with violations. Guidelines most likely to be violated included the association of beer drinking with social success and the use of content appealing to persons younger than 21 years. The alcohol industry's current self-regulatory framework is ineffective at preventing content violations but could be improved by the use of new rating procedures designed to better detect content code violations.

  5. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  6. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  7. Extracellular peptidases of the cereal pathogen Fusarium graminearum.

    Directory of Open Access Journals (Sweden)

    Rohan George Thomas Lowe

    2015-11-01

    Full Text Available The plant pathogenic fungus Fusarium graminearum (Fgr creates economic and health risks in cereals agriculture. Fgr causes head blight (or scab of wheat and stalk rot of corn, reducing yield, degrading grain quality and polluting downstream food products with mycotoxins. Fungal plant pathogens must secrete proteases to access nutrition and to breakdown the structural protein component of the plant cell wall. Research into the proteolytic activity of Fgr is hindered by the complex nature of the suite of proteases secreted. We used a systems biology approach comprising genome analysis, transcriptomics and label-free quantitative proteomics to characterise the peptidases deployed by Fgr during growth. A combined analysis of published microarray transcriptome datasets revealed seven transcriptional groupings of peptidases based on in vitro growth, in planta growth, and sporulation behaviours. An orbitrap MS/MS proteomics technique defined the extracellular proteases secreted by Fusarium graminearum. A meta-classification based on sequence characters and transcriptional/translational activity in planta and in vitro provides a platform to develop control strategies that target Fgr peptidases.

  8. Cardiac function and tadalafil used for treating fetal growth restriction in pregnant women without cardiovascular disease.

    Science.gov (United States)

    Tanaka, Kayo; Tanaka, Hiroaki; Maki, Shintaro; Kubo, Michiko; Nii, Masafumi; Magawa, Shoichi; Hatano, Fumi; Tsuji, Makoto; Osato, Kazuhiro; Kamimoto, Yuki; Umekawa, Takashi; Ikeda, Tomoaki

    2018-02-20

    The aim of the present study was to evaluate tadalafil for the treatment of fetal growth restriction (FGR) and the cardiac function in pregnant women without cardiovascular disease who used tadalafil for this reason. We examined nine pregnant women without cardiovascular disease who were using tadalafil to treat FGR. Maternal heart rate, systolic blood pressure (BP), and echocardiographic findings were assessed before and after tadalafil use. Diastolic BP was lower after compared to that before using tadalafil, but the difference was not significant. Echocardiographic findings were not significantly different before and after tadalafil use. Tadalafil did not adversely affect pregnant women without cardiovascular disease and was considered acceptable for use since it did not affect the mother's cardiac function.

  9. Development of System Based Code: Case Study of Life-Cycle Margin Evaluation

    International Nuclear Information System (INIS)

    Tai Asayama; Masaki Morishita; Masanori Tashimo

    2006-01-01

    For a leap of progress in structural deign of nuclear plant components, The late Professor Emeritus Yasuhide Asada proposed the System Based Code. The key concepts of the System Based Code are; (1) life-cycle margin optimization, (2) expansion of technical options as well as combinations of technical options beyond the current codes and standards, and (3) designing to clearly defined target reliabilities. Those concepts are very new to most of the nuclear power plant designers who are naturally obliged to design to current codes and standards; the application of the concepts of the System Based Code to design will lead to entire change of practices that designers have long been accustomed to. On the other hand, experienced designers are supposed to have expertise that can support and accelerate the development of the System Based Code. Therefore, interfacing with experienced designers is of crucial importance for the development of the System Based Code. The authors conducted a survey on the acceptability of the System Based Code concept. The results were analyzed from the possibility of improving structural design both in terms of reliability and cost effectiveness by the introduction of the System Based Code concept. It was concluded that the System Based Code is beneficial for those purposes. Also described is the expertise elicited from the results of the survey that can be reflected to the development of the System Based Code. (authors)

  10. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Control modules -- Volume 1, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Landers, N.F.; Petrie, L.M.; Knight, J.R. [Oak Ridge National Lab., TN (United States)] [and others

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. This manual is divided into three volumes: Volume 1--for the control module documentation, Volume 2--for the functional module documentation, and Volume 3 for the documentation of the data libraries and subroutine libraries.

  11. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Control modules -- Volume 1, Revision 4

    International Nuclear Information System (INIS)

    Landers, N.F.; Petrie, L.M.; Knight, J.R.

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. This manual is divided into three volumes: Volume 1--for the control module documentation, Volume 2--for the functional module documentation, and Volume 3 for the documentation of the data libraries and subroutine libraries

  12. Predictive value of mid-trimester amniotic fluid high-sensitive C-reactive protein, ferritin, and lactate dehydrogenase for fetal growth restriction

    Directory of Open Access Journals (Sweden)

    Borna Sedigheh

    2009-10-01

    Full Text Available Background: Fetal growth restriction (FGR is surprisingly common with placental dysfunction occurring in about 3% of pregnancies and despite advances in obstetric care, FGR remains a major problem in developed countries. Aim: The purpose of this study is to find out the predictive value of amniotic fluid high sensitive C-reactive protein (hs-CRP, ferritin, and lactate dehydrogenase (LDH for FGR. Materials and Methods: This prospective strategy of this study has been conducted on pregnant women who underwent genetic amniocentesis between 15th and 20th weeks of gestation. All patients were followed up on until delivery. Patients with abnormal karyotype and iatrogenic preterm delivery for fetal and maternal indications were excluded. The samples were immediately sent to laboratory for cytogenetic and biochemical examination. Non-parametric tests and receiver-operator characteristic curve analysis were used for statistical purpose. Results: A significant correlation between incremental amniotic fluid alpha fetoprotein (αFPr and LDH levels and FGR at gestational weeks 15th-20th was found out. We also found an optimum cut-off value> 140 IU/L for the amniotic fluid LDH concentration with a sensitivity of 87.5% and a specificity of 82.4% for the prediction of FGR. Conclusion: Once the LDH value is confirmed, it could serve as a prediction factor for FGR at the time of genetic amniocentesis at gestational weeks 15-20.

  13. Performance and Complexity Co-evaluation of the Advanced Video Coding Standard for Cost-Effective Multimedia Communications

    Directory of Open Access Journals (Sweden)

    Saponara Sergio

    2004-01-01

    Full Text Available The advanced video codec (AVC standard, recently defined by a joint video team (JVT of ITU-T and ISO/IEC, is introduced in this paper together with its performance and complexity co-evaluation. While the basic framework is similar to the motion-compensated hybrid scheme of previous video coding standards, additional tools improve the compression efficiency at the expense of an increased implementation cost. As a first step to bridge the gap between the algorithmic design of a complex multimedia system and its cost-effective realization, a high-level co-evaluation approach is proposed and applied to a real-life AVC design. An exhaustive analysis of the codec compression efficiency versus complexity (memory and computational costs project space is carried out at the early algorithmic design phase. If all new coding features are used, the improved AVC compression efficiency (up to 50% compared to current video coding technology comes with a complexity increase of a factor 2 for the decoder and larger than one order of magnitude for the encoder. This represents a challenge for resource-constrained multimedia systems such as wireless devices or high-volume consumer electronics. The analysis also highlights important properties of the AVC framework allowing for complexity reduction at the high system level: when combining the new coding features, the implementation complexity accumulates, while the global compression efficiency saturates. Thus, a proper use of the AVC tools maintains the same performance as the most complex configuration while considerably reducing complexity. The reported results provide inputs to assist the profile definition in the standard, highlight the AVC bottlenecks, and select optimal trade-offs between algorithmic performance and complexity.

  14. Evaluation on applicability of the rules, regulations, and industrial codes and standards for SMART development

    International Nuclear Information System (INIS)

    Choi, Suhn; Lee, C C.; Lee, C.K.; Kim, K.K.; Kim, J.P.; Kim, J.H.; Cho, B.H.; Kang, D J.; Bae, G.H.; Chung, M.; Chang, M.H.

    1999-03-01

    In this report, evaluation on applicability of the rules, regulations, and industrial codes and standards for SMART has been made. As the first step, past-to-present status of licensing structures were reviewed. Then, the rules, regulations, and standards applied to YGN 3-6 were listed and reviewed. Finally, evaluation on applicability of such rules and standards for SMART are made in each design fields. During this step technical evaluations on each items of rules, regulations and standards are made and the possible remedies or comments are suggested. The results are summarized in a tabular form and enclosed as Appendix. (Author). 8 refs., 5 tabs., 3 figs

  15. Implantation, evaluation and improvement of the diffusion code package developed by the RIS0 Research Center

    International Nuclear Information System (INIS)

    Koide, M.C.M.

    1983-01-01

    The evaluation and improvement of the diffusion code package developed by the RIS0 Research Center of Denmark have been performed. The improvements made in the package consisted in the presentation of their manuals. In order to reduce the process time of the codes an analitical boundary condition capable of representing the effects of the baffle and the reflector on the flux distribution has been calculated. Such boundary condition was obtained using a one-dimensional medium in the framework of the two group diffusion theory. The results showed that the application of this boundary condition produces very accurate results and an appreciable economy of processing time. (author) [pt

  16. Quick Response Code Secure: A Cryptographically Secure Anti-Phishing Tool for QR Code Attacks

    OpenAIRE

    Mavroeidis, Vasileios; Nicho, Mathew

    2017-01-01

    The two-dimensional quick response (QR) codes can be misleading due to the difficulty in differentiating a genuine QR code from a malicious one. Since the vulnerability is practically part of their design, scanning a malicious QR code can direct the user to cloned malicious sites resulting in revealing sensitive information. In order to evaluate the vulnerabilities and propose subsequent countermeasures, we demonstrate this type of attack through a simulated experiment, where a malicious QR c...

  17. Evaluation of the efficiency and fault density of software generated by code generators

    Science.gov (United States)

    Schreur, Barbara

    1993-01-01

    Flight computers and flight software are used for GN&C (guidance, navigation, and control), engine controllers, and avionics during missions. The software development requires the generation of a considerable amount of code. The engineers who generate the code make mistakes and the generation of a large body of code with high reliability requires considerable time. Computer-aided software engineering (CASE) tools are available which generates code automatically with inputs through graphical interfaces. These tools are referred to as code generators. In theory, code generators could write highly reliable code quickly and inexpensively. The various code generators offer different levels of reliability checking. Some check only the finished product while some allow checking of individual modules and combined sets of modules as well. Considering NASA's requirement for reliability, an in house manually generated code is needed. Furthermore, automatically generated code is reputed to be as efficient as the best manually generated code when executed. In house verification is warranted.

  18. SIMULATE-3 K coupled code applications

    Energy Technology Data Exchange (ETDEWEB)

    Joensson, Christian [Studsvik Scandpower AB, Vaesteraas (Sweden); Grandi, Gerardo; Judd, Jerry [Studsvik Scandpower Inc., Idaho Falls, ID (United States)

    2017-07-15

    This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).

  19. ASPECT: An advanced specified-profile evaluation code for tokamaks

    International Nuclear Information System (INIS)

    Stotler, D.P.; Reiersen, W.T.; Bateman, G.

    1993-03-01

    A specified-profile, global analysis code has been developed to evaluate the performance of fusion reactor designs. Both steady-state and time-dependent calculations are carried out; the results of the former can be used in defining the parameters of the latter, if desired. In the steady-state analysis, the performance is computed at a density and temperature chosen to be consistent with input limits (e.g., density and beta) of several varieties. The calculation can be made at either the intersection of the two limits or at the point of optimum performance as the density and temperature are varied along the limiting boundaries. Two measures of performance are available for this purpose: the ignition margin or the confinement level required to achieve a prescribed ignition margin. The time-dependent calculation can be configured to yield either the evolution of plasma energy as a function of time or, via an iteration scheme, the amount of auxiliary power required to achieve a desired final plasma energy

  20. Development of ADINA-J-integral code

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1988-07-01

    A general purpose finite element program ADINA (Automatic Dynamic Incremental Nonlinear Analysis), which was developed by Bathe et al., was revised to be able to calculate the J- and J-integral. This report introduced the numerical method to add this capability to the code, and the evaluation of the revised ADINA-J code by using a few of examples of the J estimation model, i.e. a compact tension specimen, a center cracked panel subjected to dynamic load, and a thick shell cylinder having inner axial crack subjected to thermal load. The evaluation testified the function of the revised code. (author)

  1. Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1991-01-01

    The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab

  2. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  3. Gap Conductance model Validation in the TASS/SMR-S code using MARS code

    International Nuclear Information System (INIS)

    Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

    2010-01-01

    Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

  4. A multiobjective approach to the genetic code adaptability problem.

    Science.gov (United States)

    de Oliveira, Lariza Laura; de Oliveira, Paulo S L; Tinós, Renato

    2015-02-19

    The organization of the canonical code has intrigued researches since it was first described. If we consider all codes mapping the 64 codes into 20 amino acids and one stop codon, there are more than 1.51×10(84) possible genetic codes. The main question related to the organization of the genetic code is why exactly the canonical code was selected among this huge number of possible genetic codes. Many researchers argue that the organization of the canonical code is a product of natural selection and that the code's robustness against mutations would support this hypothesis. In order to investigate the natural selection hypothesis, some researches employ optimization algorithms to identify regions of the genetic code space where best codes, according to a given evaluation function, can be found (engineering approach). The optimization process uses only one objective to evaluate the codes, generally based on the robustness for an amino acid property. Only one objective is also employed in the statistical approach for the comparison of the canonical code with random codes. We propose a multiobjective approach where two or more objectives are considered simultaneously to evaluate the genetic codes. In order to test our hypothesis that the multiobjective approach is useful for the analysis of the genetic code adaptability, we implemented a multiobjective optimization algorithm where two objectives are simultaneously optimized. Using as objectives the robustness against mutation with the amino acids properties polar requirement (objective 1) and robustness with respect to hydropathy index or molecular volume (objective 2), we found solutions closer to the canonical genetic code in terms of robustness, when compared with the results using only one objective reported by other authors. Using more objectives, more optimal solutions are obtained and, as a consequence, more information can be used to investigate the adaptability of the genetic code. The multiobjective approach

  5. The 1996 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1996-01-01

    The codes are named 'the Pre-processing' codes, because they are designed to pre-process ENDF/B data, for later, further processing for use in applications. This is a modular set of computer codes, each of which reads and writes evaluated nuclear data in the ENDF/B format. Each code performs one or more independent operations on the data, as described below. These codes are designed to be computer independent, and are presently operational on every type of computer from large mainframe computer to small personal computers, such as IBM-PC and Power MAC. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  6. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    Energy Technology Data Exchange (ETDEWEB)

    Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Laboratorio de Analise, Avaliacao e Gerenciamento de Risco (LabRisco/POLI/USP), São Paulo, SP (Brazil); Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: ayabe@ipen.br, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  7. Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Martins, Marcelo Ramos; Abe, Alfredo; Muniz, Rafael O.R.; Gomes, Daniel de Souza; Silva, Antonio Teixeira e

    2017-01-01

    Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes. (author)

  8. Beyond Valence and Magnitude: a Flexible Evaluative Coding System in the Brain

    Science.gov (United States)

    Gu, Ruolei; Lei, Zhihui; Broster, Lucas; Wu, Tingting; Jiang, Yang; Luo, Yue-jia

    2013-01-01

    Outcome evaluation is a cognitive process that plays an important role in our daily lives. In most paradigms utilized in the field of experimental psychology, outcome valence and outcome magnitude are the two major features investigated. The classical “independent coding model” suggest that outcome valence and outcome magnitude are evaluated by separate neural mechanisms that may be mapped onto discrete event-related potential (ERP) components: feedback-related negativity (FRN) and the P3, respectively. To examine this model, we presented outcome valence and magnitude sequentially rather than simultaneously. The results reveal that when only outcome valence or magnitude is known, both the FRN and the P3 encode that outcome feature; when both aspects of outcome are known, the cognitive functions of the two components dissociate: the FRN responds to the information available in the current context, while the P3 pattern depends on outcome presentation sequence. The current study indicates that the human evaluative system, indexed in part by the FRN and the P3, is more flexible than previous theories suggested. PMID:22019775

  9. Effect of flue gas recirculation during oxy-fuel combustion in a rotary cement kiln

    International Nuclear Information System (INIS)

    Granados, David A.; Chejne, Farid; Mejía, Juan M.; Gómez, Carlos A.; Berrío, Ariel; Jurado, William J.

    2014-01-01

    The effect of Flue Gas Recirculation (FGR) during Oxy-Fuel Combustion in a Rotary Cement Kiln was analyzed by using a CFD model applied to coal combustion process. The CFD model is based on 3D-balance equations for mass, species, energy and momentum. Turbulence and radiation model coupled to a chemical kinetic mechanism for pyrolysis processes, gas–solid and gas–gas reactions was included to predicts species and flame temperature distribution, as well as convective and radiation energy fluxes. The model was used to study coal combustion with air and with oxygen for FGR between 30 and 85% as controller parameter for temperature in the process. Flame length effect and heat transfer by convection and radiation to the clinkering process for several recirculation ratios was studied. Theoretical studies predicted a located increase of energy flux and a reduction in flame length with respect to the traditional system which is based on air combustion. The impact of FGR on the oxy-fuel combustion process and different energy scenarios in cement kilns to increase energy efficiency and clinker production were studied and evaluated. Simulation results were in close agreement with experimental data, where the maximum deviation was 7%

  10. An Empirical Evaluation of the US Beer Institute’s Self-Regulation Code Governing the Content of Beer Advertising

    Science.gov (United States)

    Xuan, Ziming; Damon, Donna; Noel, Jonathan

    2013-01-01

    Objectives. We evaluated advertising code violations using the US Beer Institute guidelines for responsible advertising. Methods. We applied the Delphi rating technique to all beer ads (n = 289) broadcast in national markets between 1999 and 2008 during the National Collegiate Athletic Association basketball tournament games. Fifteen public health professionals completed ratings using quantitative scales measuring the content of alcohol advertisements (e.g., perceived actor age, portrayal of excessive drinking) according to 1997 and 2006 versions of the Beer Institute Code. Results. Depending on the code version, exclusion criteria, and scoring method, expert raters found that between 35% and 74% of the ads had code violations. There were significant differences among producers in the frequency with which ads with violations were broadcast, but not in the proportions of unique ads with violations. Guidelines most likely to be violated included the association of beer drinking with social success and the use of content appealing to persons younger than 21 years. Conclusions. The alcohol industry’s current self-regulatory framework is ineffective at preventing content violations but could be improved by the use of new rating procedures designed to better detect content code violations. PMID:23947318

  11. A code guidance system for integrated nuclear data evaluation system on the basis of knowledge engineering technology

    International Nuclear Information System (INIS)

    Fukahori, Tokio; Nakagawa, Tsuneo

    1994-01-01

    The integrated nuclear data evaluation system (INDES) is being made in order to support the nuclear data evaluation work. A guidance system in INDES, 'Evaluation Tutor (ET)', is under development in order to support users in selecting the most suitable set of theoretical calculation codes by applying knowledge engineering technology and the experiences of evaluation work for JENDL-3. In this paper, the function of ET is introduced as well as the functions and databases of INDES. An example run of ET for 56 Fe in the 1-20 MeV neutron energy region is also explained. (author)

  12. Sildenafil Citrate Increases Fetal Weight in a Mouse Model of Fetal Growth Restriction with a Normal Vascular Phenotype

    Science.gov (United States)

    Dilworth, Mark Robert; Andersson, Irene; Renshall, Lewis James; Cowley, Elizabeth; Baker, Philip; Greenwood, Susan; Sibley, Colin Peter; Wareing, Mark

    2013-01-01

    Fetal growth restriction (FGR) is defined as the inability of a fetus to achieve its genetic growth potential and is associated with a significantly increased risk of morbidity and mortality. Clinically, FGR is diagnosed as a fetus falling below the 5th centile of customised growth charts. Sildenafil citrate (SC, Viagra™), a potent and selective phosphodiesterase-5 inhibitor, corrects ex vivo placental vascular dysfunction in FGR, demonstrating potential as a therapy for this condition. However, many FGR cases present without an abnormal vascular phenotype, as assessed by Doppler measures of uterine/umbilical artery blood flow velocity. Thus, we hypothesized that SC would not increase fetal growth in a mouse model of FGR, the placental-specific Igf2 knockout mouse, which has altered placental exchange capacity but normal placental blood flow. Fetal weights were increased (by 8%) in P0 mice following maternal SC treatment (0.4 mg/ml) via drinking water. There was also a trend towards increased placental weight in treated P0 mice (P = 0.056). Additionally, 75% of the P0 fetal weights were below the 5th centile, the criterion used to define human FGR, of the non-treated WT fetal weights; this was reduced to 51% when dams were treated with SC. Umbilical artery and vein blood flow velocity measures confirmed the lack of an abnormal vascular phenotype in the P0 mouse; and were unaffected by SC treatment. 14C-methylaminoisobutyric acid transfer (measured to assess effects on placental nutrient transporter activity) per g placenta was unaffected by SC, versus untreated, though total transfer was increased, commensurate with the trend towards larger placentas in this group. These data suggest that SC may improve fetal growth even in the absence of an abnormal placental blood flow, potentially affording use in multiple sub-populations of individuals presenting with FGR. PMID:24204949

  13. Sildenafil citrate increases fetal weight in a mouse model of fetal growth restriction with a normal vascular phenotype.

    Directory of Open Access Journals (Sweden)

    Mark Robert Dilworth

    Full Text Available Fetal growth restriction (FGR is defined as the inability of a fetus to achieve its genetic growth potential and is associated with a significantly increased risk of morbidity and mortality. Clinically, FGR is diagnosed as a fetus falling below the 5(th centile of customised growth charts. Sildenafil citrate (SC, Viagra™, a potent and selective phosphodiesterase-5 inhibitor, corrects ex vivo placental vascular dysfunction in FGR, demonstrating potential as a therapy for this condition. However, many FGR cases present without an abnormal vascular phenotype, as assessed by Doppler measures of uterine/umbilical artery blood flow velocity. Thus, we hypothesized that SC would not increase fetal growth in a mouse model of FGR, the placental-specific Igf2 knockout mouse, which has altered placental exchange capacity but normal placental blood flow. Fetal weights were increased (by 8% in P0 mice following maternal SC treatment (0.4 mg/ml via drinking water. There was also a trend towards increased placental weight in treated P0 mice (P = 0.056. Additionally, 75% of the P0 fetal weights were below the 5(th centile, the criterion used to define human FGR, of the non-treated WT fetal weights; this was reduced to 51% when dams were treated with SC. Umbilical artery and vein blood flow velocity measures confirmed the lack of an abnormal vascular phenotype in the P0 mouse; and were unaffected by SC treatment. (14C-methylaminoisobutyric acid transfer (measured to assess effects on placental nutrient transporter activity per g placenta was unaffected by SC, versus untreated, though total transfer was increased, commensurate with the trend towards larger placentas in this group. These data suggest that SC may improve fetal growth even in the absence of an abnormal placental blood flow, potentially affording use in multiple sub-populations of individuals presenting with FGR.

  14. In vivo placental MRI shape and textural features predict fetal growth restriction and postnatal outcome.

    Science.gov (United States)

    Dahdouh, Sonia; Andescavage, Nickie; Yewale, Sayali; Yarish, Alexa; Lanham, Diane; Bulas, Dorothy; du Plessis, Adre J; Limperopoulos, Catherine

    2018-02-01

    To investigate the ability of three-dimensional (3D) MRI placental shape and textural features to predict fetal growth restriction (FGR) and birth weight (BW) for both healthy and FGR fetuses. We recruited two groups of pregnant volunteers between 18 and 39 weeks of gestation; 46 healthy subjects and 34 FGR. Both groups underwent fetal MR imaging on a 1.5 Tesla GE scanner using an eight-channel receiver coil. We acquired T2-weighted images on either the coronal or the axial plane to obtain MR volumes with a slice thickness of either 4 or 8 mm covering the full placenta. Placental shape features (volume, thickness, elongation) were combined with textural features; first order textural features (mean, variance, kurtosis, and skewness of placental gray levels), as well as, textural features computed on the gray level co-occurrence and run-length matrices characterizing placental homogeneity, symmetry, and coarseness. The features were used in two machine learning frameworks to predict FGR and BW. The proposed machine-learning based method using shape and textural features identified FGR pregnancies with 86% accuracy, 77% precision and 86% recall. BW estimations were 0.3 ± 13.4% (mean percentage error ± standard error) for healthy fetuses and -2.6 ± 15.9% for FGR. The proposed FGR identification and BW estimation methods using in utero placental shape and textural features computed on 3D MR images demonstrated high accuracy in our healthy and high-risk cohorts. Future studies to assess the evolution of each feature with regard to placental development are currently underway. 2 Technical Efficacy: Stage 2 J. Magn. Reson. Imaging 2018;47:449-458. © 2017 International Society for Magnetic Resonance in Medicine.

  15. LMWH in the prevention of preeclampsia and fetal growth restriction in women without thrombophilia. A systematic review and meta-analysis.

    Science.gov (United States)

    Mastrolia, Salvatore Andrea; Novack, Lena; Thachil, Jecko; Rabinovich, Anat; Pikovsky, Oleg; Klaitman, Vered; Loverro, Giuseppe; Erez, Offer

    2016-10-28

    Placental mediated pregnancy complications such as preeclampsia and fetal growth restriction (FGR) are common, serious, and associated with increased morbidity and mortality. We conducted a systematic review and meta-analysis to determine the effect of treatment with low-molecular-weight heparins (LMWHs) for secondary prevention of these complications in non thrombophilic women. We searched the electronic databases PubMed, Scopus, and Cochrane Library for randomised controlled trials addressing this question. Five studies including 403 patients met the inclusion criteria, 68 developed preeclampsia and 118 FGR. The studies were very heterogeneous in terms of inclusion criteria, LMWH preparation, and dosage. Meta-analyses were performed using random-effect models. The overall use of LMWHs was associated with a risk reduction for preeclampsia (Relative risk (RR) 0.366; 95 % confidence interval (CI), 0.219-0.614) and FGR (RR 0.409; 95 % CI, 0.195-0.932) vs. no treatment. From the data available for analysis it appears that the use of Dalteparin is associated with a risk reduction for preeclampsia (p=0.002) and FGR (ppreeclampsia (p=0.013) but not for FGR (p=0.3). In spite of the small number of studies addressing the research question, and the high variability among them, our meta-analysis found a modest beneficial effect of LMWH for secondary prevention of preeclampsia and FGR. Further studies are needed to address these questions before a definite conclusion can be reached.

  16. ENDF utility codes version 6.8

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1992-01-01

    Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-5 and ENDF-6. Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting codes PLOTEF, GRALIB, graphic device interface subroutine library INTLIB; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package which is designed for CDC, IBM, DEC and PC computers, can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

  17. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey.

    Science.gov (United States)

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospital grounds. These QR code bracelets link to the QR Code Identity website, where detailed information is stored; a smartphone or standalone QR code scanner can be used to scan the code. The design of this system allows authorized personnel (e.g., paramedics, firefighters, or police) to access more detailed patient information than the average smartphone user: emergency service professionals are authorized to access patient medical histories to improve the accuracy of medical treatment. In Istanbul, we tested the self-designed system with 174 participants. To analyze the QR Code Identity Tag system's usability, the participants completed the System Usability Scale questionnaire after using the system.

  18. Decorin expression is decreased in first trimester placental tissue from pregnancies with small for gestation age infants at birth

    NARCIS (Netherlands)

    Murthi, P.; van Zanten, D. E.; Eijsink, J. J. H.; Borg, A. J.; Stevenson, J. L.; Kalionis, B.; Chui, A. K.; Said, J. M.; Brennecke, S. P.; Emrich, J. J. H. M.

    Fetal growth restriction (FGR) is a leading cause of perinatal morbidity and mortality. FGR pregnancies are often associated with histological evidence of placental vascular thrombosis. The proteoglycans are important components and regulators of vascular homeostasis. Previous studies from our

  19. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Crecy, F. de; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  20. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  1. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  2. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Crecy, F de; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  3. Structural evaluation method for class 1 vessels by using elastic-plastic finite element analysis in code case of JSME rules on design and construction

    International Nuclear Information System (INIS)

    Asada, Seiji; Hirano, Takashi; Nagata, Tetsuya; Kasahara, Naoto

    2008-01-01

    A structural evaluation method by using elastic-plastic finite element analysis has been developed and published as a code case of Rules on Design and Construction for Nuclear Power Plants (The First Part: Light Water Reactor Structural Design Standard) in the JSME Codes for Nuclear Power Generation Facilities. Its title is 'Alternative Structural Evaluation Criteria for Class 1 Vessels Based on Elastic-Plastic Finite Element Analysis' (NC-CC-005). This code case applies elastic-plastic analysis to evaluation of such failure modes as plastic collapse, thermal ratchet, fatigue and so on. Advantage of this evaluation method is free from stress classification, consistently use of Mises stress and applicability to complex 3-dimensional structures which are hard to be treated by the conventional stress classification method. The evaluation method for plastic collapse has such variation as the Lower Bound Approach Method, Twice-Elastic-Slope Method and Elastic Compensation Method. Cyclic Yield Area (CYA) based on elastic analysis is applied to screening evaluation of thermal ratchet instead of secondary stress evaluation, and elastic-plastic analysis is performed when the CYA screening criteria is not satisfied. Strain concentration factors can be directly calculated based on elastic-plastic analysis. (author)

  4. On-line monitoring and inservice inspection in codes

    International Nuclear Information System (INIS)

    Bartonicek, J.; Zaiss, W.; Bath, H.R.

    1999-01-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

  5. Demonstration study on shielding safety analysis code (VI)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering

    1999-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)

  6. A Monte Carlo computer code for evaluating energy loss of 10 keV to 10 MeV ions in amorphous silicon materials

    International Nuclear Information System (INIS)

    Erramli, H.; Elbounagui, O.; Misdaq, M.A.; Merzouki, A.

    2007-01-01

    The basic concepts of a computer simulation code for determining the energy loss of ions in the 10 keV to 10 MeV energy range in amorphous silicon materials were presented and discussed. Data obtained were found in good agreement with those obtained by using a SRIM programme. Electronic and nuclear energy losses were evaluated. Variation of the energy loss as a function of the incident ion energy were studied. This new computer code is a good tool for evaluating stopping powers of various materials for light and heavy ions

  7. Requests from use experience of ORIGEN code. Activity of the working group on evaluation of nuclide generation and depletion

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo

    2005-01-01

    A questionnaire survey was carried out through the committee members of the working group on evaluation of nuclide generation and depletion about the demand accuracy of the ORIGEN code which is used widely in various fields of design analysis and evaluation. WG committee asked each organization's ORIGEN user, and obtained the replay from various fields. (author)

  8. Specialized Monte Carlo codes versus general-purpose Monte Carlo codes

    International Nuclear Information System (INIS)

    Moskvin, Vadim; DesRosiers, Colleen; Papiez, Lech; Lu, Xiaoyi

    2002-01-01

    The possibilities of Monte Carlo modeling for dose calculations and optimization treatment are quite limited in radiation oncology applications. The main reason is that the Monte Carlo technique for dose calculations is time consuming while treatment planning may require hundreds of possible cases of dose simulations to be evaluated for dose optimization. The second reason is that general-purpose codes widely used in practice, require an experienced user to customize them for calculations. This paper discusses the concept of Monte Carlo code design that can avoid the main problems that are preventing wide spread use of this simulation technique in medical physics. (authors)

  9. Validation study of computer code SPHINCS for sodium fire safety evaluation of fast reactor

    International Nuclear Information System (INIS)

    Yamaguchi, Akira; Tajima, Yuji

    2003-01-01

    A computer code SPHINCS solves coupled phenomena of thermal hydraulics and sodium fire based on a multi-zone model. It deals with an arbitrary number of rooms, each of which is connected mutually by doorways and penetrations. With regard to the combustion phenomena, a flame sheet model and a liquid droplet combustion model are used for pool and spray fires, respectively, with the chemical equilibrium model based on the Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermalhydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The authors analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS code has been validated with regard to pool combustion phenomena. Further research needs are identified for pool spreading modeling considering thermal deformation of steel liner and measurement of pool fluidity property as a mixture of liquid sodium and reaction products. The SPHINCS code is to be used mainly in the safety evaluation of the consequence of a sodium fire accident in a liquid metal cooled fast reactor as well as fire safety analysis in general

  10. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  11. Consensus definition and essential reporting parameters of selective fetal growth restriction in twin pregnancy: a Delphi procedure

    NARCIS (Netherlands)

    Khalil, Asma; Beune, Irene; Hecher, Kurt; Wynia, Klaske; Ganzevoort, Wessel; Reed, Keith; Lewi, Liesbeth; Oepkes, Dick; Gratacos, Eduardo; Thilaganathan, Basky; Gordijn, Sanne J.

    2018-01-01

    Twin pregnancies complicated by selective fetal growth restriction (sFGR) are associated with increased perinatal mortality and morbidity. Inconsistences in the diagnostic criteria for sFGR employed in existing studies hinder the ability to compare or combine their findings. It is therefore

  12. Dynamic conservation of forest genetic resources in 33 European countries

    NARCIS (Netherlands)

    Lefevre, F.; Koskela, J.; Hubert, J.; Kraigher, H.; Longauer, R.; Olrik, D.C.; Vries, de S.M.G.

    2013-01-01

    Dynamic conservation of forest genetic resources (FGR) means maintaining the genetic diversity of trees within an evolutionary process and allowing generation turnover in the forest. We assessed the network of forests areas managed for the dynamic conservation of FGR (conservation units) across

  13. Expression of Biglycan in First Trimester Chorionic Villous Sampling Placental Samples and Altered Function in Telomerase-Immortalized Microvascular Endothelial Cells

    NARCIS (Netherlands)

    Chui, Amy; Gunatillake, Tilini; Brennecke, Shaun P.; Ignjatovic, Vera; Monagle, Paul T.; Whitelock, John M.; van Zanten, Dagmar E.; Eijsink, Jasper; Wang, Yao; Deane, James; Borg, Anthony J.; Stevenson, Janet; Erwich, Jan Jaap; Said, Joanne M.; Murthi, Padma

    Objective-Biglycan (BGN) has reduced expression in placentae from pregnancies complicated by fetal growth restriction (FGR). We used first trimester placental samples from pregnancies with later small for gestational age (SGA) infants as a surrogate for FGR. The functional consequences of reduced

  14. Relationship between general movements in neonates who were growth restricted in utero and prenatal Doppler flow patterns

    NARCIS (Netherlands)

    Tanis, J. C.; Schmitz, D. M.; Boelen, M. R.; Casarella, L.; Berg, van den Paul; Bilardo, C. M.; Bos, A. F.

    2016-01-01

    Objective To investigate whether Doppler pulsatility indices (PIs) of the fetal circulation in cases of fetal growth restriction (FGR) are associated with the general movements (GMs) of the neonate after birth. Methods This was a prospective observational cohort study including pregnancies with FGR

  15. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.; De Crecy, F.

    1994-01-01

    This paper deals with the application of the adjoint sensitivity method (ASM) to thermal hydraulic codes. The advantage of the method is to use small central processing unit time in comparison with the usual approach requiring one complete code run per sensitivity determination. In the first part the mathematical aspects of the problem are treated, and the applicability of the method of the functional-type response of a thermal hydraulic model is demonstrated. On a simple example of non-linear hyperbolic equation (Burgers equation) the problem has been analysed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the continuous ASM and the discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the discrete ASM constitutes a practical solution for thermal hydraulic codes. The application of the discrete ASM to the thermal hydraulic safety code CATHARE is then presented for two examples. They demonstrate that the discrete ASM constitutes an efficient tool for the analysis of code sensitivity. ((orig.))

  16. Evaluation and implementation of QR Code Identity Tag system for Healthcare in Turkey

    OpenAIRE

    Uzun, Vassilya; Bilgin, Sami

    2016-01-01

    For this study, we designed a QR Code Identity Tag system to integrate into the Turkish healthcare system. This system provides QR code-based medical identification alerts and an in-hospital patient identification system. Every member of the medical system is assigned a unique QR Code Tag; to facilitate medical identification alerts, the QR Code Identity Tag can be worn as a bracelet or necklace or carried as an ID card. Patients must always possess the QR Code Identity bracelets within hospi...

  17. Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

    Directory of Open Access Journals (Sweden)

    Kim Kyung-O

    2016-01-01

    Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

  18. Current status of high energy nucleon-meson transport code

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)

  19. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  20. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  1. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  2. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  3. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  4. Induction technology optimization code

    International Nuclear Information System (INIS)

    Caporaso, G.J.; Brooks, A.L.; Kirbie, H.C.

    1992-01-01

    A code has been developed to evaluate relative costs of induction accelerator driver systems for relativistic klystrons. The code incorporates beam generation, transport and pulsed power system constraints to provide an integrated design tool. The code generates an injector/accelerator combination which satisfies the top level requirements and all system constraints once a small number of design choices have been specified (rise time of the injector voltage and aspect ratio of the ferrite induction cores, for example). The code calculates dimensions of accelerator mechanical assemblies and values of all electrical components. Cost factors for machined parts, raw materials and components are applied to yield a total system cost. These costs are then plotted as a function of the two design choices to enable selection of an optimum design based on various criteria. (Author) 11 refs., 3 figs

  5. Iterative nonlinear unfolding code: TWOGO

    International Nuclear Information System (INIS)

    Hajnal, F.

    1981-03-01

    a new iterative unfolding code, TWOGO, was developed to analyze Bonner sphere neutron measurements. The code includes two different unfolding schemes which alternate on successive iterations. The iterative process can be terminated either when the ratio of the coefficient of variations in terms of the measured and calculated responses is unity, or when the percentage difference between the measured and evaluated sphere responses is less than the average measurement error. The code was extensively tested with various known spectra and real multisphere neutron measurements which were performed inside the containments of pressurized water reactors

  6. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules, F9-F11

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes.

  7. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules, F9-F11

    International Nuclear Information System (INIS)

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with three of the functional modules in the code. Those are the Morse-SGC for the SCALE system, Heating 7.2, and KENO V.a. The manual describes the latest released versions of the codes

  8. Neutronic evolution of SENA reactor during the first and second cycles. Comparison between the experimental power distributions obtained from the in-core instrumentation evaluation code CIRCE and the theoretical power values computed with the two-dimensional diffusion-evolution code EVOE

    International Nuclear Information System (INIS)

    Andrieux, Chantal

    1976-03-01

    The neutronic evolution of the reacteur Sena during the first and second cycles is presented. The experimental power distributions, obtained from the in-core instrumentation evaluation code CIRCE are compared with the theoretical powers calculated with the two-dimensional diffusion-evolution code EVOE. The CIRCE code allows: the study of the evolution of the principal parameters of the core, the comparison of the results of measured and theoretical estimates. Therefore this study has a great interest for the knowledge of the neutronic evolution of the core, as well as the validation of the refinement of theoretical estimation methods. The core calculation methods and requisite data for the evaluation of the measurements are presented after a brief description of the SENA core and its inner instrumentation. The principle of the in-core instrumentation evaluation code CIRCE, and calculation of the experimental power distributions and nuclear core parameters are then exposed. The results of the evaluation are discussed, with a comparison of the theoretical and experimental results. Taking account of the approximations used, these results, as far as the first and second cycles at SENA are concerned, are satisfactory, the deviations between theoretical and experimental power distributions being lower than 3% at the middle of the reactor and 9% at the periphery [fr

  9. Validation uncertainty of MATRA code for subchannel void distributions

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae-Hyun; Kim, S. J.; Kwon, H.; Seo, K. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To extend code capability to the whole core subchannel analysis, pre-conditioned Krylov matrix solvers such as BiCGSTAB and GMRES are implemented in MATRA code as well as parallel computing algorithms using MPI and OPENMP. It is coded by fortran 90, and has some user friendly features such as graphic user interface. MATRA code was approved by Korean regulation body for design calculation of integral-type PWR named SMART. The major role subchannel code is to evaluate core thermal margin through the hot channel analysis and uncertainty evaluation for CHF predictions. In addition, it is potentially used for the best estimation of core thermal hydraulic field by incorporating into multiphysics and/or multi-scale code systems. In this study we examined a validation process for the subchannel code MATRA specifically in the prediction of subchannel void distributions. The primary objective of validation is to estimate a range within which the simulation modeling error lies. The experimental data for subchannel void distributions at steady state and transient conditions was provided on the framework of OECD/NEA UAM benchmark program. The validation uncertainty of MATRA code was evaluated for a specific experimental condition by comparing the simulation result and experimental data. A validation process should be preceded by code and solution verification. However, quantification of verification uncertainty was not addressed in this study. The validation uncertainty of the MATRA code for predicting subchannel void distribution was evaluated for a single data point of void fraction measurement at a 5x5 PWR test bundle on the framework of OECD UAM benchmark program. The validation standard uncertainties were evaluated as 4.2%, 3.9%, and 2.8% with the Monte-Carlo approach at the axial levels of 2216 mm, 2669 mm, and 3177 mm, respectively. The sensitivity coefficient approach revealed similar results of uncertainties but did not account for the nonlinear effects on the

  10. New evaluated neutron cross section libraries for the GEANT4 code

    International Nuclear Information System (INIS)

    Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.

    2012-04-01

    The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)

  11. Verification of reactor safety codes

    International Nuclear Information System (INIS)

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  12. Demonstration study on shielding safety analysis code (8)

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan)

    2001-03-01

    Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this study is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated. (1) A {sup 3}He detector and some instruments are added to the former detection system to increase the detection sensitivity in pulsed neutron measurements. Using the new detection system, the skyshine of neutrons from 45 MeV LINAC facility are measured in the distance up to 350 m. (2) To estimate the spectrum of leakage neutron from the facility, {sup 3}He detector with moderators is constructed and the response functions of the detector are calculated using the MCNP simulation code. The leakage spectrum in the facility are measured and unfolded using the SAND-II code. (3) Using the EGS code and/or MCNP code, neutron yields by the photo-nuclear reaction in the lead target are calculated. Then, the neutron fluence at some points including the duct (from which neutrons leaks and is considered to be a skyshine source) is simulated by MCNP MONTE CARLO code. (4) In the distance up to 350 m from the facility, neutron fluence due to the skyshine process are calculated and compared with the experimental results. The comparison gives a fairly good agreement. (author)

  13. Placental oxidative stress and maternal endothelial function in pregnant women with normotensive fetal growth restriction.

    Science.gov (United States)

    Yoshida, Atsumi; Watanabe, Kazushi; Iwasaki, Ai; Kimura, Chiharu; Matsushita, Hiroshi; Wakatsuki, Akihiko

    2018-04-01

    The purpose of this study was to investigate the relationship between placental oxidative stress and maternal endothelial function in pregnant women with normotensive fetal growth restriction (FGR). We examined serum concentrations of oxygen free radicals (d-ROMs), maternal angiogenic factor (PlGF), and sFlt-1, placental oxidative DNA damage, and maternal endothelial function in 17 women with early-onset preeclampsia (PE), 18 with late-onset PE, 14 with normotensive FGR, and 21 controls. Flow-mediated vasodilation (FMD) was assessed as a marker of maternal endothelial function. Immunohistochemical analysis was performed to measure the proportion of placental trophoblast cell nuclei staining positive for 8-hydroxy-2'-deoxyguanosine (8-OHdG), a marker of oxidative DNA damage. Maternal serum d-ROM, sFlt-1 concentrations, and FMD did not significantly differ between the control and normotensive FGR groups. The proportion of nuclei staining positive for 8-OHdG was significantly higher in the normotensive FGR group relative to the control group. Our findings demonstrate that, despite the presence of placental oxidative DNA damage as observed in PE patients, pregnant women with normotensive FGR show no increase in the concentrations of sFlt-1 and d-ROMs, or a decrease in FMD.

  14. Establishment of a JSME code for the evaluation of high-cycle thermal fatigue in mixing tees

    International Nuclear Information System (INIS)

    Moriya, Shoichi; Fukuda, Toshihiko; Matsunaga, Tomoya; Hirayama, Hiroshi; Shiina, Kouji; Tanimoto, Koichi

    2004-01-01

    This paper describes a JSME code for high-cycle thermal fatigue evaluation by thermal striping in mixing tees with hot and cold water flows. The evaluation of thermal striping in a mixing tee has four steps to screen design parameters one-by-one according to the severity of the thermal load assessed from design conditions using several evaluation charts. In order to make these charts, visualization tests with acrylic pipes and temperature measurement tests with metal pipes were conducted. The influence of the configurations of mixing tees, flow velocity ratio, pipe diameter ratio and so on was examined from the results of the experiments. This paper makes a short mention of the process of providing these charts. (author)

  15. Preparation of the TRANSURANUS code for TEMELIN NPP

    International Nuclear Information System (INIS)

    Klouzal, J.

    2011-01-01

    Since 2010 Temelin NPP started using TVSA-T fuel supplied by JSC TVEL. The transition process included implementation of several new core reload design codes. TRANSURANUS code was selected for the evaluation of the fuel rod thermomechanical performance. The adaptation and validation of the code was performed by Nuclear Research Institute Rez. TRANSURANUS code contains wide selection of alternative models for most of phenomena important for the fuel behaviour. It was therefore necessary to select, based on a comparison with experimental data, those most suitable for the modeling of TVSA-T fuel rods. In some cases, new models were implemented. Software tools and methodology for the evaluation of the proposed core reload design using TRANSURANUS code were also developed in NRI. The software tools include the interface to core physics code ANDREA and a set of scripts for an automated execution and processing of the computational runs. Independent confirmation of some of the vendor specified core reload design criteria was performed using TRANSURANUS. (authors)

  16. Computation of the bounce-average code

    International Nuclear Information System (INIS)

    Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.

    1977-01-01

    The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended

  17. A development of computer code for evaluating internal radiation dose through ingestion and inhalation pathways

    International Nuclear Information System (INIS)

    Lee, Jeong Ho; Lee, Chang Woo; Choi, Yong Ho; Chun, Ki Jung; Kim, Kook Chan; Kim, Sang Bok; Kim, Jin Kyu

    1991-07-01

    The computer codes were developed to evaluate internal radiation dose when radioactive isotopes released from nuclear facilities are taken through ingestion and inhalation pathways. Food chain models and relevant data base representing the agricultural and social environment of Korea are set up. An equilibrium model-KFOOD, which can deal with routine releases from a nuclear facility and a dynamic model-ECOREA, which is suitable for the description of acute radioactivity release following nuclear accident. (Author)

  18. Evaluation of angular integrals in the generation of transfer matrices for multigroup transport codes

    International Nuclear Information System (INIS)

    Garcia, R.D.M.

    1985-01-01

    The generalization of a semi-analytical technique for the evaluation of angular integrals that appear in the generation of elastic and discrete inelastic tranfer matrices for transport codes is carried out. In contrast to the generalized series expansions which are found to be too complex and thus of little practical value, when compared to the Gaussian quadrature technique, the recursion relations developed in this work are superior to the quadrature scheme, for those cases where the round-off error propagation is not significant. (Author) [pt

  19. EVALUATION OF USAGE AND APPLICATION AREAS OF QR CODES IN SERVICE INDUSTRY

    Directory of Open Access Journals (Sweden)

    Aysel SANAL

    2018-01-01

    Full Text Available The use of QR codes accelerates the sharing of information and provides more practical access to information. In today's information age, in the limited area unlimited information, data and contents can be transferred with using QR code. This study examines how consumers use QR code technology using by the service sector and aim of inform consumers about their perception and usage levels. In the application part of the study, 180 consumers responded the survey questions. The t-test, ANOVA, variance analysis and regression analysis method were used to test hypotheses established in the research. Thus, factors affecting the perceptions of consumers on QR code technology have been identified. The finance and banking sectors have been identified as the sectors in which consumers use the QR code most frequently, and the speed and availability factors for this sector have been analyzed separately.

  20. A comparison of oxide thickness predictability from the perspective of codes

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo-Young; Shin, Hye-In; Kim, Kyung-Tae; Han, Hee-Tak; Kim, Hong-Jin; Kim, Yong-Hwan [KEPCO Nuclear Fuel Co. Ltd., Daejeon (Korea, Republic of)

    2016-10-15

    In Korea, OPR1000 and Westinghouse type nuclear power plant reactor fuel rods oxide thickness has been evaluated by imported code A. Because of this, there have been multiple constraints in operation and maintenance of fuel rod design system. For this reason, there has been a growing demand to establish an independent fuel rod design system. To meet this goal, KNF has recently developed its own code B for fuel rod design. The objective of this study is to compare oxide thickness prediction performance between code A and code B and to check the validity of predicting corrosion behaviors of newly developed code B. This study is based on Pool Side Examination (PSE) data for the performance confirmation. For the examination procedures, the oxide thickness measurement methods and equipment of PSE are described in detail. In this study, code B is confirmed conservatism and validity on evaluating cladding oxide thickness through the comparison with code A. Code prediction values show higher value than measured data from PSE. Throughout this study, the values by code B are evaluated and proved to be valid in a view point of the oxide thickness evaluation. However, the code B input for prediction has been made by designer's judgment with complex handwork that might be lead to excessive conservative result and ineffective design process with some possibility of errors.

  1. AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

    It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

  2. BER EVALUATION OF LDPC CODES WITH GMSK IN NAKAGAMI FADING CHANNEL

    Directory of Open Access Journals (Sweden)

    Surbhi Sharma

    2010-06-01

    Full Text Available LDPC codes (Low Density Parity Check Codes have already proved its efficacy while showing its performance near to the Shannon limit. Channel coding schemes are spectrally inefficient as using an unfiltered binary data stream to modulate an RF carrier that will produce an RF spectrum of considerable bandwidth. Techniques have been developed to improve this bandwidth inefficiency or spectral efficiency, and ease detection. GMSK or Gaussian-filtered Minimum Shift Keying uses a Gaussian Filter of an appropriate bandwidth so as to make system spectrally efficient. A Nakagami model provides a better explanation to less and more severe conditions than the Rayleigh and Rician model and provide a better fit to the mobile communication channel data. In this paper we have demonstrated the performance of Low Density Parity Check codes with GMSK modulation (BT product=0.25 technique in Nakagami fading channel. In results it is shown that average bit error rate decreases as the ‘m’ parameter increases (Less fading.

  3. Application of CATE 2.0 code on evaluating activated corrosion products in a PWR cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Jingyu; Li, Lu; Chen, Yixue [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering

    2017-03-15

    In PWR plants, most Occupational Radiation Exposure (ORE) for personnel results from Activated Corrosion Products (ACPs) in the cooling loop. In order to evaluate the ACPs in the cooling loop, a three-region transport model is built up based on the theory of driving force from the concentration difference in CATE 2.0 code. In order to analyze the nuclide composition of ACPs, the EAF-2007 nuclear database is embedded in CATE 2.0. The case of MIT PCCL test loop is simulated to test the availability of CATE 2.0 on PWR ACPs evaluation, and the activity of Co-58 and Co-60 after operation for 42 days calculated by CATE 2.0 is consistent with that from the code CRUDSIM adopted by MIT. Then, the nuclide composition of ACPs is analyzed in detail respectively for operation of 42 days and 12 months using CATE 2.0. The results show that the short-lived nuclides contribute a majority of the activity in the regions of in-flux wall and coolant, while the long-lived nuclides contribute most of the activity in the region of out-flux wall.

  4. Isolation of basal membrane proteins from BeWo cells and their expression in placentas from fetal growth-restricted pregnancies.

    Science.gov (United States)

    Oh, Soo-Young; Hwang, Jae Ryoung; Lee, Yoonna; Choi, Suk-Joo; Kim, Jung-Sun; Kim, Jong-Hwa; Sadovsky, Yoel; Roh, Cheong-Rae

    2016-03-01

    The syncytiotrophoblast, a key barrier between the mother and fetus, is a polarized epithelium composed of a microvillus and basal membrane (BM). We sought to characterize BM proteins of BeWo cells in relation to hypoxia and to investigate their expression in placentas from pregnancies complicated by fetal growth restriction (FGR). We isolated the BM fraction of BeWo cells by the cationic colloidal silica method and identified proteins enriched in this fraction by mass spectrometry. We evaluated the effect of hypoxia on the expression and intracellular localization of identified proteins and compared their expression in BM fractions of FGR placentas to those from normal pregnancies. We identified BM proteins from BeWo cells. Among BM proteins, we further characterized heme oxygenase-1 (HO-1), voltage-dependent anion channel-1 (VDAC1), and ribophorin II (RPN2), based on their relevance to placental biology. Hypoxia enhanced the localization of these proteins to the BM of BeWo cells. HO-1, VDAC1, and RPN2 were selectively expressed in the human placental BM fraction. C-terminally truncated HO-1 was identified in placental BM fractions, and its BM expression was significantly reduced in FGR placentas than in normal placentas. Interestingly, a truncated HO-1 construct was predominantly localized in the BM in response to hypoxia and co-localized with VDAC1 in BeWo cells. Hypoxia increased the BM localization of HO-1, VDAC1, and RPN2 proteins. FGR significantly reduced the expression of truncated HO-1, which was surmised to co-localize with VDAC1 in hypoxic BeWo cells. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Control modules C4, C6

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U. S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume is part of the manual related to the control modules for the newest updated version of this computational package.

  6. Reliability of coded data to identify earliest indications of cognitive decline, cognitive evaluation and Alzheimer's disease diagnosis: a pilot study in England.

    Science.gov (United States)

    Dell'Agnello, Grazia; Desai, Urvi; Kirson, Noam Y; Wen, Jody; Meiselbach, Mark K; Reed, Catherine C; Belger, Mark; Lenox-Smith, Alan; Martinez, Carlos; Rasmussen, Jill

    2018-03-22

    Evaluate the reliability of using diagnosis codes and prescription data to identify the timing of symptomatic onset, cognitive assessment and diagnosis of Alzheimer's disease (AD) among patients diagnosed with AD. This was a retrospective cohort study using the UK Clinical Practice Research Datalink (CPRD). The study cohort consisted of a random sample of 50 patients with first AD diagnosis in 2010-2013. Additionally, patients were required to have a valid text-field code and a hospital episode or a referral in the 3 years before the first AD diagnosis. The earliest indications of cognitive impairment, cognitive assessment and AD diagnosis were identified using two approaches: (1) using an algorithm based on diagnostic codes and prescription drug information and (2) using information compiled from manual review of both text-based and coded data. The reliability of the code-based algorithm for identifying the earliest dates of the three measures described earlier was evaluated relative to the comprehensive second approach. Additionally, common cognitive assessments (with and without results) were described for both approaches. The two approaches identified the same first dates of cognitive symptoms in 33 (66%) of the 50 patients, first cognitive assessment in 29 (58%) patients and first AD diagnosis in 43 (86%) patients. Allowing for the dates from the two approaches to be within 30 days, the code-based algorithm's success rates increased to 74%, 70% and 94%, respectively. Mini-Mental State Examination was the most commonly observed cognitive assessment in both approaches; however, of the 53 tests performed, only 19 results were observed in the coded data. The code-based algorithm shows promise for identifying the first AD diagnosis. However, the reliability of using coded data to identify earliest indications of cognitive impairment and cognitive assessments is questionable. Additionally, CPRD is not a recommended data source to identify results of cognitive

  7. Survey of computer codes applicable to waste facility performance evaluations

    International Nuclear Information System (INIS)

    Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

    1988-01-01

    This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

  8. Quality assurance aspects of the computer code CODAR2

    International Nuclear Information System (INIS)

    Maul, P.R.

    1986-03-01

    The computer code CODAR2 was developed originally for use in connection with the Sizewell Public Inquiry to evaluate the radiological impact of routine discharges to the sea from the proposed PWR. It has subsequently bee used to evaluate discharges from Heysham 2. The code was frozen in September 1983, and this note gives details of its verification, validation and evaluation. Areas where either improved modelling methods or more up-to-date information relevant to CODAR2 data bases have subsequently become available are indicated; these will be incorporated in any future versions of the code. (author)

  9. Tokamak plasma power balance calculation code (TPC code) outline and operation manual

    International Nuclear Information System (INIS)

    Fujieda, Hirobumi; Murakami, Yoshiki; Sugihara, Masayoshi.

    1992-11-01

    This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

  10. Recent improvements in modelling fission gas release and rod deformation on metallic fuel in LMR

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung-Oon; Kim, Young Jin

    2000-01-01

    Metallic fuel design is a key feature to assure LMR core safety goals. To date, a large effort has been devoted to the development of the MACSIS code for metallic fuel rod design and the evaluation of operational limits under irradiation conditions. The updated models of fission gas release, fuel core swelling, and rod deformation are incorporated into the correspondence routines in MACSIS MOD1. The MACSIS MOD1 which is a new version of MACSIS, has been partly benchmarked on FGR, fuel swelling and rod deformation comparing with the results of U-Zr and U-Pu-Zr metal fuels irradiated in LMRs. The MACSIS MOD1 predicts, relatively well, the absolute magnitudes and trends of the gas release and rod deformations depending on burn-up, and it gives better agreement with the experimental data than the previous predictions of MACSIS and the results of the empirical model

  11. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  12. The 1992 ENDF Pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1992-01-01

    This document summarizes the 1992 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. Included are the codes CONVERT, MERGER, LINEAR, RECENT, SIGMA1, LEGEND, FIXUP, GROUPIE, DICTION, MIXER, VIRGIN, COMPLOT, EVALPLOT, RELABEL. Some of the functions of these codes are: to calculate cross-sections from resonance parameters; to calculate angular distributions, group average, mixtures of cross-sections, etc; to produce graphical plottings and data comparisons. The codes are designed to operate on virtually any type of computer including PC's. They are available from the IAEA Nuclear Data Section, free of charge upon request, on magnetic tape or a set of HD diskettes. (author)

  13. OPAL reactor calculations using the Monte Carlo code serpent

    Energy Technology Data Exchange (ETDEWEB)

    Ferraro, Diego; Villarino, Eduardo [Nuclear Engineering Dept., INVAP S.E., Rio Negro (Argentina)

    2012-03-15

    In the present work the Monte Carlo cell code developed by VTT Serpent v1.1.14 is used to model the MTR fuel assemblies (FA) and control rods (CR) from OPAL (Open Pool Australian Light-water) reactor in order to obtain few-group constants with burnup dependence to be used in the already developed reactor core models. These core calculations are performed using CITVAP 3-D diffusion code, which is well-known reactor code based on CITATION. Subsequently the results are compared with those obtained by the deterministic calculation line used by INVAP, which uses the Collision Probability Condor cell-code to obtain few-group constants. Finally the results are compared with the experimental data obtained from the reactor information for several operation cycles. As a result several evaluations are performed, including a code to code cell comparison at cell and core level and calculation-experiment comparison at core level in order to evaluate the Serpent code actual capabilities. (author)

  14. Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1983-06-01

    The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

  15. Double-digit coding of examination math problems

    Directory of Open Access Journals (Sweden)

    Agnieszka Sułowska

    2013-09-01

    Full Text Available Various methods are used worldwide to evaluate student solutions to examination tasks. Usually the results simply provide information about student competency and after aggregation, are also used as a tool of making comparisons between schools. In particular, the standard evaluation methods do not allow conclusions to be drawn about possible improvements of teaching methods. There are however, task assessment methods which not only allow description of student achievement, but also possible causes of failure. One such method, which can be applied to extended response tasks, is double-digit coding which has been used in some international educational research. This paper presents the first Polish experiences of applying this method to examination tasks in mathematics, using a special coding key to carry out the evaluation. Lessons learned during the coding key construction and its application in the assessment process are described.

  16. Evaluation of Extended CCSDS Reed-Solomon Codes for Bandwidth efficiency

    DEFF Research Database (Denmark)

    Andersen, Jakob Dahl; Justesen, Jørn; Larsen, Knud J.

    1999-01-01

    The present CCSDS recommendation for Telemetry Channel Coding was originally written around twenty years ago. The appearance of the Turbo coding scheme has made an inclusion of this powerful scheme desirable, and thus it becomes natural also to perform a major rewriting of the other part of the r....... Finally, we present advantages and disadvantages by placing the frame synchronizer before and after the Viterbi decoder, and we suggest an option where the attached sync marker is not convolutionally encoded....

  17. A compendium of computer codes in fault tree analysis

    International Nuclear Information System (INIS)

    Lydell, B.

    1981-03-01

    In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)

  18. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  19. SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP

    International Nuclear Information System (INIS)

    Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi

    2009-05-01

    Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)

  20. Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1984-12-01

    The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

  1. Joint Source-Channel Coding by Means of an Oversampled Filter Bank Code

    Directory of Open Access Journals (Sweden)

    Marinkovic Slavica

    2006-01-01

    Full Text Available Quantized frame expansions based on block transforms and oversampled filter banks (OFBs have been considered recently as joint source-channel codes (JSCCs for erasure and error-resilient signal transmission over noisy channels. In this paper, we consider a coding chain involving an OFB-based signal decomposition followed by scalar quantization and a variable-length code (VLC or a fixed-length code (FLC. This paper first examines the problem of channel error localization and correction in quantized OFB signal expansions. The error localization problem is treated as an -ary hypothesis testing problem. The likelihood values are derived from the joint pdf of the syndrome vectors under various hypotheses of impulse noise positions, and in a number of consecutive windows of the received samples. The error amplitudes are then estimated by solving the syndrome equations in the least-square sense. The message signal is reconstructed from the corrected received signal by a pseudoinverse receiver. We then improve the error localization procedure by introducing a per-symbol reliability information in the hypothesis testing procedure of the OFB syndrome decoder. The per-symbol reliability information is produced by the soft-input soft-output (SISO VLC/FLC decoders. This leads to the design of an iterative algorithm for joint decoding of an FLC and an OFB code. The performance of the algorithms developed is evaluated in a wavelet-based image coding system.

  2. Coding visual features extracted from video sequences.

    Science.gov (United States)

    Baroffio, Luca; Cesana, Matteo; Redondi, Alessandro; Tagliasacchi, Marco; Tubaro, Stefano

    2014-05-01

    Visual features are successfully exploited in several applications (e.g., visual search, object recognition and tracking, etc.) due to their ability to efficiently represent image content. Several visual analysis tasks require features to be transmitted over a bandwidth-limited network, thus calling for coding techniques to reduce the required bit budget, while attaining a target level of efficiency. In this paper, we propose, for the first time, a coding architecture designed for local features (e.g., SIFT, SURF) extracted from video sequences. To achieve high coding efficiency, we exploit both spatial and temporal redundancy by means of intraframe and interframe coding modes. In addition, we propose a coding mode decision based on rate-distortion optimization. The proposed coding scheme can be conveniently adopted to implement the analyze-then-compress (ATC) paradigm in the context of visual sensor networks. That is, sets of visual features are extracted from video frames, encoded at remote nodes, and finally transmitted to a central controller that performs visual analysis. This is in contrast to the traditional compress-then-analyze (CTA) paradigm, in which video sequences acquired at a node are compressed and then sent to a central unit for further processing. In this paper, we compare these coding paradigms using metrics that are routinely adopted to evaluate the suitability of visual features in the context of content-based retrieval, object recognition, and tracking. Experimental results demonstrate that, thanks to the significant coding gains achieved by the proposed coding scheme, ATC outperforms CTA with respect to all evaluation metrics.

  3. An EG-VEGF-dependent decrease in homeobox gene NKX3.1 contributes to cytotrophoblast dysfunction: a possible mechanism in human fetal growth restriction.

    Science.gov (United States)

    Murthi, P; Brouillet, S; Pratt, A; Borg, Aj; Kalionis, B; Goffin, F; Tsatsaris, V; Munaut, C; Feige, Jj; Benharouga, M; Fournier, T; Alfaidy, N

    2015-07-21

    Idiopathic fetal growth restriction (FGR) is frequently associated with placental insufficiency. Previous reports have provided evidence that EG-VEGF (endocrine gland derived-vascular endothelial growth factor), a placental secreted protein, is expressed during the first trimester of pregnancy, controls both trophoblast proliferation and invasion, and its increased expression is associated with human FGR. In this study, we hypothesise that EG-VEGF-dependent change in placental homeobox gene expressions contribute to trophoblast dysfunction in idiopathic FGR. The changes in EG-VEGF-dependent homeobox gene expressions were determined using a Homeobox gene cDNA array on placental explants of 8-12 weeks' gestation after stimulation with EG-VEGF in vitro for 24 hours. The Homeobox gene array identified a >5-fold increase in HOXA9, HOXC8, HOXC10, HOXD1, HOXD8, HOXD9 and HOXD11, while NKX 3.1 showed a >2 fold-decrease in mRNA expression compared to untreated controls. Homeobox gene NKX3.1 was selected as a candidate because it is a downstream target of EG-VEGF and its expression and functional role are largely unknown in control and idiopathic FGR-affected placentae. Real-time PCR and immunoblotting showed a significant decrease in NKX3.1 mRNA and protein levels, respectively, in placentae from FGR compared to control pregnancies. Gene inactivation in vitro using short-interference RNA specific for NKX3.1 demonstrated an increase in BeWo cell differentiation and a decrease in HTR8-SVneo proliferation. We conclude that the decreased expression of homeobox gene NKX3.1 down-stream of EG-VEGF may contribute to the trophoblast dysfunction associated with idiopathic FGR pregnancies.

  4. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  5. Evaluations on power ramp data of PWR fuels by FROST and THERMOST codes

    International Nuclear Information System (INIS)

    Murai, K.; Ogawa, S.; Nuno, H.; Kondo, Y.

    1987-01-01

    An evaluation is presented of power ramp data of Mitsubishi's PWR fuel rods tested in R-2, Studsvik, which was analysed by FROST and THERMOST codes. The analyses give good predictions for measured diameter changes and on-power rod elongations. The work indicates that FROST is capable of analysing both radial and axial pellet-cladding mechanism interaction (PCMI) appropriately, and that predicted states of PCMI (i.e. stress and strain which cannot be measured directly) are considered to be reliable. The ramp data used in the present analyses were obtained in two joint programmes with five Japanese PWR utilities (KEPCO, KYEPCO, SEPCO, HEPCO, and JAPCO). (UK)

  6. Codeword Structure Analysis for LDPC Convolutional Codes

    Directory of Open Access Journals (Sweden)

    Hua Zhou

    2015-12-01

    Full Text Available The codewords of a low-density parity-check (LDPC convolutional code (LDPC-CC are characterised into structured and non-structured. The number of the structured codewords is dominated by the size of the polynomial syndrome former matrix H T ( D , while the number of the non-structured ones depends on the particular monomials or polynomials in H T ( D . By evaluating the relationship of the codewords between the mother code and its super codes, the low weight non-structured codewords in the super codes can be eliminated by appropriately choosing the monomials or polynomials in H T ( D , resulting in improved distance spectrum of the mother code.

  7. Development of simplified decommissioning cost estimation code for nuclear facilities

    International Nuclear Information System (INIS)

    Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu

    2010-01-01

    The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report. (author)

  8. An evaluation of the background introduced from the coded aperture mask in the low energy gamma-ray telescope ZEBRA

    International Nuclear Information System (INIS)

    Butler, R.C.; Caroli, E.; Di Cocco, G.; Maggioli, P.P.; Spizzichino, A.; Charalambous, P.M.; Dean, A.J.; Drane, M.; Gil, A.; Stephen, J.B.; Perotti, F.; Villa, G.; Badiali, M.; La Padula, C.; Polcaro, F.; Ubertini, P.

    1984-01-01

    The background which arises from the presence of a coded aperture mask is evaluated. The major contributions which have been considered here are the interactions with the mask of the isotropic gamma-ray background, a parallel gamma-ray beam, neutrons and the effect of the mask element profile. It is shown that none of these factors conbribute to a significant excess or modulation in the background counting rate over the detection plane. In this way the use of a passive rather than an active coded aperture mask is seen to be suitable for use in a low energy gamma-ray telescope. (orig.)

  9. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

    International Nuclear Information System (INIS)

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE

  10. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE.

  11. Consistent Code Qualification Process and Application to WWER-1000 NPP

    International Nuclear Information System (INIS)

    Berthon, A.; Petruzzi, A.; Giannotti, W.; D'Auria, F.; Reventos, F.

    2006-01-01

    Calculation analysis by application of the system codes are performed to evaluate the NPP or the facility behavior during a postulated transient or to evaluate the code capability. The calculation analysis constitutes a process that involves the code itself, the data of the reference plant, the data about the transient, the nodalization, and the user. All these elements affect one each other and affect the results. A major issue in the use of mathematical model is constituted by the model capability to reproduce the plant or facility behavior under steady state and transient conditions. These aspects constitute two main checks that must be satisfied during the qualification process. The first of them is related to the realization of a scheme of the reference plant; the second one is related to the capability to reproduce the transient behavior. The aim of this paper is to describe the UMAE (Uncertainty Method based on Accuracy Extrapolation) methodology developed at University of Pisa for qualifying a nodalization and analysing the calculated results and to perform the uncertainty evaluation of the system code by the CIAU code (Code with the capability of Internal Assessment of Uncertainty). The activity consists with the re-analysis of the Experiment BL-44 (SBLOCA) performed in the LOBI facility and the analysis of a Kv-scaling calculation of the WWER-1000 NPP nodalization taking as reference the test BL-44. Relap5/Mod3.3 has been used as thermal-hydraulic system code and the standard procedure adopted at University of Pisa has been applied to show the capability of the code to predict the significant aspects of the transient and to obtain a qualified nodalization of the WWER-1000 through a systematic qualitative and quantitative accuracy evaluation. The qualitative accuracy evaluation is based on the selection of Relevant Thermal-hydraulic Aspects (RTAs) and is a prerequisite to the application of the Fast Fourier Transform Based Method (FFTBM) which quantifies

  12. Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

    International Nuclear Information System (INIS)

    Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

    2006-01-01

    as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined. (authors)

  13. Sample problem manual for benchmarking of cask analysis codes

    International Nuclear Information System (INIS)

    Glass, R.E.

    1988-02-01

    A series of problems have been defined to evaluate structural and thermal codes. These problems were designed to simulate the hypothetical accident conditions given in Title 10 of the Code of Federal Regulation, Part 71 (10CFR71) while retaining simple geometries. This produced a problem set that exercises the ability of the codes to model pertinent physical phenomena without requiring extensive use of computer resources. The solutions that are presented are consensus solutions based on computer analyses done by both national laboratories and industry in the United States, United Kingdom, France, Italy, Sweden, and Japan. The intent of this manual is to provide code users with a set of standard structural and thermal problems and solutions which can be used to evaluate individual codes. 19 refs., 19 figs., 14 tabs

  14. [Is DRG Coding too Important to be Left to Physicians? - Evaluation of Economic Efficiency by Health Economists in a University Medical Centre].

    Science.gov (United States)

    Burger, F; Walgenbach, M; Göbel, P; Parbs, S; Neugebauer, E

    2017-04-01

    Background: We investigated and evaluated the cost effectiveness of coding by health care economists in a centre for orthopaedics and trauma surgery in Germany, by quantifying and comparing the financial efficiency of physicians with basic knowledge of the DRG-system with the results of healthcare economists with in-depth knowledge (M.Sc.). In addition, a hospital survey was performed to establish how DRG-coding is being performed and the identity of the persons involved. Material and Methods: In a prospective and controlled study, 200 in-patients were coded by a healthcare economist (study group). Prior to that, the same cases were coded by physicians with basic training in the DRG-system, who made up the control group. All cases were picked randomly and blinded without informing the physicians coding the controls, in order to avoid any Hawthorne effect. We evaluated and measured the effective weighting within the G-DRG, the DRG returns per patient, the overall DRG return, and the additional time needed. For the survey, questionnaires were sent to 1200 German hospitals. The completed questionnaire was analysed using a statistical program. Results: The return difference per patient between controls and the study group was significantly greater (2472 ± 337 €; p DRG case reports was 1277 (2500-62,300). Coding was performed in 69 % of cases by doctors, 19 % by skilled specialists for DRG coding and in 8 % together. Overall satisfaction with the DRG was described by 61 % of respondents as good or excellent. Conclusion: Our prospective and controlled study quantifies the cost efficiency of health economists in a centre of orthopaedics and trauma surgery in Germany for the first time. We provide some initial evidence that health economists can enhance the CMI, the resulting DRG return per patient as well as the overall DRG return. Data from the survey shows that in many hospitals there is great reluctance to leave the coding to specialists only. Georg

  15. Empirical Evaluation of Superposition Coded Multicasting for Scalable Video

    KAUST Repository

    Chun Pong Lau; Shihada, Basem; Pin-Han Ho

    2013-01-01

    In this paper we investigate cross-layer superposition coded multicast (SCM). Previous studies have proven its effectiveness in exploiting better channel capacity and service granularities via both analytical and simulation approaches. However

  16. GARLIC — A general purpose atmospheric radiative transfer line-by-line infrared-microwave code: Implementation and evaluation

    International Nuclear Information System (INIS)

    Schreier, Franz; Gimeno García, Sebastián; Hedelt, Pascal; Hess, Michael; Mendrok, Jana; Vasquez, Mayte; Xu, Jian

    2014-01-01

    A suite of programs for high resolution infrared-microwave atmospheric radiative transfer modeling has been developed with emphasis on efficient and reliable numerical algorithms and a modular approach appropriate for simulation and/or retrieval in a variety of applications. The Generic Atmospheric Radiation Line-by-line Infrared Code — GARLIC — is suitable for arbitrary observation geometry, instrumental field-of-view, and line shape. The core of GARLIC's subroutines constitutes the basis of forward models used to implement inversion codes to retrieve atmospheric state parameters from limb and nadir sounding instruments. This paper briefly introduces the physical and mathematical basics of GARLIC and its descendants and continues with an in-depth presentation of various implementation aspects: An optimized Voigt function algorithm combined with a two-grid approach is used to accelerate the line-by-line modeling of molecular cross sections; various quadrature methods are implemented to evaluate the Schwarzschild and Beer integrals; and Jacobians, i.e. derivatives with respect to the unknowns of the atmospheric inverse problem, are implemented by means of automatic differentiation. For an assessment of GARLIC's performance, a comparison of the quadrature methods for solution of the path integral is provided. Verification and validation are demonstrated using intercomparisons with other line-by-line codes and comparisons of synthetic spectra with spectra observed on Earth and from Venus. - Highlights: • High resolution infrared-microwave radiative transfer model. • Discussion of algorithmic and computational aspects. • Jacobians by automatic/algorithmic differentiation. • Performance evaluation by intercomparisons, verification, validation

  17. Toric Varieties and Codes, Error-correcting Codes, Quantum Codes, Secret Sharing and Decoding

    DEFF Research Database (Denmark)

    Hansen, Johan Peder

    We present toric varieties and associated toric codes and their decoding. Toric codes are applied to construct Linear Secret Sharing Schemes (LSSS) with strong multiplication by the Massey construction. Asymmetric Quantum Codes are obtained from toric codes by the A.R. Calderbank P.W. Shor and A.......M. Steane construction of stabilizer codes (CSS) from linear codes containing their dual codes....

  18. New Predictive Model at 11+0 to 13+6 Gestational Weeks for Early-Onset Preeclampsia With Fetal Growth Restriction.

    Science.gov (United States)

    Chang, Ying; Chen, Xu; Cui, Hong-Yan; Li, Xing; Xu, Ya-Ling

    2017-05-01

    The aim of the present study was to determine a predictive model for early-onset preeclampsia with fetal growth restriction (FGR) to be used at 11 +0 to 13 +6 gestational weeks, by combining the maternal serum level of pregnancy-associated plasma protein-A (PAPP-A), placental growth factor (PLGF), placental protein 13 (PP13), soluble endoglin (sEng), mean arterial pressure (MAP), and uterine artery Doppler. This was a retrospective cohort study of 4453 pregnant women. Uterine artery Doppler examination was conducted in the first trimester. Maternal serum PAPP-A, PLGF, PP13, and sEng were measured. Mean arterial pressure was obtained. Women were classified as with/without early-onset preeclampsia, and women with preeclampsia were classified as with/without FGR. Receiver operating characteristic analysis was performed to determine the value of the model. There were 30 and 32 pregnant women with early-onset preeclampsia with and without FGR. The diagnosis rate of early-onset preeclampsia with FGR was 67.4% using the predictive model when the false positive rate was set at 5% and 73.2% when the false positive rate was 10%. The predictive model (MAP, uterine artery Doppler measurements, and serum biomarkers) had some predictive value for the early diagnosis (11 +0 to 13 +6 gestational weeks) of early-onset preeclampsia with FGR.

  19. Correlation of VCAM-1 expression in serum, cord blood, and placental tissue with gestational hypertension associated with fetal growth restriction in women from Xingtai Hebei, China.

    Science.gov (United States)

    Zhang, H G; Guo, W; Gu, H F; Chen, S B; Wang, J Q; Qiao, Z X; Ma, H S; Geng, S X

    2016-08-26

    The aim of this study was to investigate the expression of vascular adhesion molecule (VCAM)-1 in the maternal serum, cord blood, and placental tissue of pregnant women from Xingtai, Hebei, with gestational hypertension (GH) combined with fetal growth restriction (FGR). A total of 108 patients with GH combined with FGR (GH-FGR), 60 patients with GH alone (GH), and 50 healthy pregnant women (control) were recruited to this study. VCAM- 1 expression was detected in the maternal serum and cord blood by enzyme-linked immunosorbent assay, and in the placental tissue by immunohistochemistry. VCAM-1 expression was significantly higher in the maternal serum of patients with GH-FGR (164.38 ± 60.35) and GH alone (103.85 ± 54.47) than in the serum of the control population (46.70 ± 21.79; P 0.05). Moreover, the VCAM-1 expression rates were significantly higher and lower in the vascular endothelial and trophoblastic cells of the placenta of patients with GH-FGR (74.71 and 56.1%) and GH (72.98 and 55.36%), respectively, compared to those in the control subjects (46.48 and 95.11%). Therefore, we concluded that VCAM- 1 plays an important role in the development and generation of GH. Additionally, the low VCAM-1 expression in the trophoblastic cell could be correlated to the pathogenesis and progression of GH.

  20. Evaluation of linear heat rates for the power-to-melt tests on 'JOYO' using the Monte-Carlo code 'MVP'

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto

    2000-04-01

    The linear heat rates of the power-to-melt (PTM) tests, performed with B5D-1 and B5D-2 subassemblies on the Experimental Fast Reactor 'JOYO', are evaluated with the continuous energy Monte-Carlo code, MVP. We can apply a whole core model to MVP, but it takes very long time for the calculation. Therefore, judging from the structure of B5D subassembly, we used the MVP code to calculate the radial distribution of linear heat rate and used the deterministic method to calculate the axial distribution. We also derived the formulas for this method. Furthermore, we evaluated the error of the linear heat rate, by evaluating the experimental error of the reactor power, the statistical error of Monte-Carlo method, the calculational model error of the deterministic method and so on. On the other hand, we also evaluated the burnup rate of the B5D assembly and compared with the measured value in the post-irradiation test. The main results are following: B5D-1 (B5101, F613632, core center). Linear heat rate: 600 W/cm±2.2%. Burnup rate: 0.977. B5D-2 (B5214, G80124, core center). Linear heat rate: 641 W/cm±2.2%. Burnup rate: 0.886. (author)

  1. Development and evaluation of a portable CZT coded aperture gamma-camera

    Energy Technology Data Exchange (ETDEWEB)

    Montemont, G.; Monnet, O.; Stanchina, S.; Maingault, L.; Verger, L. [CEA, LETI, Minatec Campus, Univ. Grenoble Alpes, 38054 Grenoble, (France); Carrel, F.; Lemaire, H.; Schoepff, V. [CEA, LIST, 91191 Gif-sur-Yvette, (France); Ferrand, G.; Lalleman, A.-S. [CEA, DAM, DIF, 91297 Arpajon, (France)

    2015-07-01

    We present the design and the evaluation of a CdZnTe (CZT) based gamma camera using a coded aperture mask. This camera, based on a 8 cm{sup 3} detection module, is small enough to be portable and battery-powered (4 kg weight and 4 W power dissipation). As the detector has spectral capabilities, the gamma camera allows isotope identification and colored imaging, by affecting one color channel to each identified isotope. As all data processing is done at real time, the user can directly observe the outcome of an acquisition and can immediately react to what he sees. We first present the architecture of the system, how the detector works, and its performances. After, we focus on the imaging technique used and its strengths and limitations. Finally, results concerning sensitivity, spatial resolution, field of view and multi-isotope imaging are shown and discussed. (authors)

  2. Development and evaluation of a portable CZT coded aperture gamma-camera

    International Nuclear Information System (INIS)

    Montemont, G.; Monnet, O.; Stanchina, S.; Maingault, L.; Verger, L.; Carrel, F.; Lemaire, H.; Schoepff, V.; Ferrand, G.; Lalleman, A.-S.

    2015-01-01

    We present the design and the evaluation of a CdZnTe (CZT) based gamma camera using a coded aperture mask. This camera, based on a 8 cm 3 detection module, is small enough to be portable and battery-powered (4 kg weight and 4 W power dissipation). As the detector has spectral capabilities, the gamma camera allows isotope identification and colored imaging, by affecting one color channel to each identified isotope. As all data processing is done at real time, the user can directly observe the outcome of an acquisition and can immediately react to what he sees. We first present the architecture of the system, how the detector works, and its performances. After, we focus on the imaging technique used and its strengths and limitations. Finally, results concerning sensitivity, spatial resolution, field of view and multi-isotope imaging are shown and discussed. (authors)

  3. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  4. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  5. Contributions to the validation of the ASTEC V1 code

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

    2004-01-01

    In the frame of PHEBEN2 project (Validation of the severe accidents codes for applications to nuclear power plants, based on the PHEBUS FP experiments), a project developed within the EU research Frame Program 5 (FP5), the INR-Pitesti's team has received the task of determining the ASTEC code sensitivity. The PHEBEN2 project has been initiated in 1998 and gathered 13 partners from 6 EU member states. To the project 4 partners from 3 candidate states (Hungary, Bulgaria and Romania) joined later. The works were contracted with the European Commission (under FIKS-CT1999-00009 contract) that supports financially the research effort up to about 50%. According to the contract provisions, INR's team participated in developing the Working Package 1 (WP1) which refers to validation of the integral computation codes that use the PHOEBUS experimental data and the Working Package 3 (WP3) referring to the evaluation of the codes to be applied in nuclear power plants for risk evaluation, nuclear safety margin evaluation and determination/evaluation of the measures to be adopted in case of severe accident. The present work continues the efforts to validate preliminarily the ASTEC code. Focused are the the stand-alone sensitivity analyses applied to two most important modules of the code, namely DIVA and SOPHAEROS

  6. Development of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations

    Energy Technology Data Exchange (ETDEWEB)

    Ohnuki, Akira; Akimoto, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kamo, Hideki

    1996-11-01

    In order to perform design calculations for a passive safety reactor with good accuracy by a multidimensional two-fluid model, we developed an analysis code, ACE-3D, which can apply for evaluation of constitutive equations. The developed code has the following features: 1. The basic equations are based on 3-dimensional two-fluid model and the orthogonal or the cylindrical coordinate system can be selected. The fluid system is air-water or steam-water. 2. The basic equations are formulated by the finite-difference scheme of staggered mesh. The convection term is formulated by an upwind scheme and the diffusion term by a center-difference scheme. 3. Semi-implicit numerical scheme is adopted and the mass and the energy equations are treated equally in convergent steps for Jacobi equations. 4. The interfacial stress term consists of drag force, life force, turbulent dispersion force, wall force and virtual mass force. 5. A {kappa}-{epsilon} turbulent model for bubbly flow is incorporated as the turbulent model. The predictive capability of ACE-3D has been verified using a data-base for bubbly flow in a small-scale vertical pipe. In future, the constitutive equations will be improved with a data-base in a large vertical pipe developed in our laboratory and we have a plan to construct a reliable analytical tool through the improvement work, the progress of calculational speed with vector and parallel processing, the assessments for phase change terms and so on. This report describes the outline for the basic equations and the finite-difference equations in ACE-3D code and also the outline for the program structure. Besides, the results for the assessments of ACE-3D code for the small-scale pipe are summarized. (author)

  7. Development of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Akimoto, Hajime; Kamo, Hideki.

    1996-11-01

    In order to perform design calculations for a passive safety reactor with good accuracy by a multidimensional two-fluid model, we developed an analysis code, ACE-3D, which can apply for evaluation of constitutive equations. The developed code has the following features: 1. The basic equations are based on 3-dimensional two-fluid model and the orthogonal or the cylindrical coordinate system can be selected. The fluid system is air-water or steam-water. 2. The basic equations are formulated by the finite-difference scheme of staggered mesh. The convection term is formulated by an upwind scheme and the diffusion term by a center-difference scheme. 3. Semi-implicit numerical scheme is adopted and the mass and the energy equations are treated equally in convergent steps for Jacobi equations. 4. The interfacial stress term consists of drag force, life force, turbulent dispersion force, wall force and virtual mass force. 5. A κ-ε turbulent model for bubbly flow is incorporated as the turbulent model. The predictive capability of ACE-3D has been verified using a data-base for bubbly flow in a small-scale vertical pipe. In future, the constitutive equations will be improved with a data-base in a large vertical pipe developed in our laboratory and we have a plan to construct a reliable analytical tool through the improvement work, the progress of calculational speed with vector and parallel processing, the assessments for phase change terms and so on. This report describes the outline for the basic equations and the finite-difference equations in ACE-3D code and also the outline for the program structure. Besides, the results for the assessments of ACE-3D code for the small-scale pipe are summarized. (author)

  8. Geographic Information Systems using CODES linked data (Crash outcome data evaluation system)

    Science.gov (United States)

    2001-04-01

    This report presents information about geographic information systems (GIS) and CODES linked data. Section one provides an overview of a GIS and the benefits of linking to CODES. Section two outlines the basic issues relative to the types of map data...

  9. Modelling of Rod No 8 in IFA-597:3

    International Nuclear Information System (INIS)

    Malen, K.

    2002-06-01

    A Westinghouse Atom 8x8 fuel rod irradiated in the Ringhals 1 BWR for 12 years to a local burnup of about 67 MWd/kgU was refabricated, instrumented with centreline thermocouple and pressure transducer, and irradiated in IFA-597.2 for about 20 days and in IFA-597.3 for about four months. The rod was then sent to Kjeller for puncturing and then to the Studsvik hot cells for detailed post-irradiation examinations. The peak centreline, temperature was close to 1350 deg C. The total fission gas release (FGR) determined from the puncturing was approximately 20 %. Electron probe microanalysis on a fuel section from the central part of the rod showed that virtually 100 % Xe release had occurred in the central part of the pellet out to about half the pellet radius, and this thermal release from the central part of the fuel accounted for the measured total FGR. Optical and scanning electron microscopy of the fuel cross-section showed complete pellet-clad bonding as well as an extensive high burnup 'rim' structure extending at least 0,15 mm in from the fuel surface. The fuel microstructure was characterised at different radial positions in the pellet. This report describes modelling of the rod behaviour using the code SKIROD, in particular fuel temperature and fission gas release. The transient response of the fuel centre line temperature after a scram is also modelled using the code TOODEE2. The modelling results are compared to the experimental results

  10. Quantifying reactor safety margins: Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology

    International Nuclear Information System (INIS)

    Boyack, B.E.; Duffey, R.B.; Griffith, P.

    1988-01-01

    In August 1988, the Nuclear Regulatory Commission (NRC) approved the final version of a revised rule on the acceptance of emergency core cooling systems (ECCS) entitled ''Emergency Core Cooling System; Revisions to Acceptance Criteria.'' The revised rule states an alternate ECCS performance analysis, based on best-estimate methods, may be used to provide more realistic estimates of plant safety margins, provided the licensee quantifies the uncertainty of the estimates and included that uncertainty when comparing the calculated results with prescribed acceptance limits. To support the revised ECCS rule, the NRC and its contractors and consultants have developed and demonstrated a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology. It is an auditable, traceable, and practical method for combining quantitative analyses and expert opinions to arrive at computed values of uncertainty. This paper provides an overview of the CSAU evaluation methodology and its application to a postulated cold-leg, large-break loss-of-coolant accident in a Westinghouse four-loop pressurized water reactor with 17 /times/ 17 fuel. The code selected for this demonstration of the CSAU methodology was TRAC-PF1/MOD1, Version 14.3. 23 refs., 5 figs., 1 tab

  11. Coding in pigeons: Multiple-coding versus single-code/default strategies.

    Science.gov (United States)

    Pinto, Carlos; Machado, Armando

    2015-05-01

    To investigate the coding strategies that pigeons may use in a temporal discrimination tasks, pigeons were trained on a matching-to-sample procedure with three sample durations (2s, 6s and 18s) and two comparisons (red and green hues). One comparison was correct following 2-s samples and the other was correct following both 6-s and 18-s samples. Tests were then run to contrast the predictions of two hypotheses concerning the pigeons' coding strategies, the multiple-coding and the single-code/default. According to the multiple-coding hypothesis, three response rules are acquired, one for each sample. According to the single-code/default hypothesis, only two response rules are acquired, one for the 2-s sample and a "default" rule for any other duration. In retention interval tests, pigeons preferred the "default" key, a result predicted by the single-code/default hypothesis. In no-sample tests, pigeons preferred the key associated with the 2-s sample, a result predicted by multiple-coding. Finally, in generalization tests, when the sample duration equaled 3.5s, the geometric mean of 2s and 6s, pigeons preferred the key associated with the 6-s and 18-s samples, a result predicted by the single-code/default hypothesis. The pattern of results suggests the need for models that take into account multiple sources of stimulus control. © Society for the Experimental Analysis of Behavior.

  12. Verification of SACI-2 computer code comparing with experimental results of BIBLIS-A and LOOP-7 computer code

    International Nuclear Information System (INIS)

    Soares, P.A.; Sirimarco, L.F.

    1984-01-01

    SACI-2 is a computer code created to study the dynamic behaviour of a PWR nuclear power plant. To evaluate the quality of its results, SACI-2 was used to recalculate commissioning tests done in BIBLIS-A nuclear power plant and to calculate postulated transients for Angra-2 reactor. The results of SACI-2 computer code from BIBLIS-A showed as much good agreement as those calculated with the KWU Loop 7 computer code for Angra-2. (E.G.) [pt

  13. Performance measures for transform data coding.

    Science.gov (United States)

    Pearl, J.; Andrews, H. C.; Pratt, W. K.

    1972-01-01

    This paper develops performance criteria for evaluating transform data coding schemes under computational constraints. Computational constraints that conform with the proposed basis-restricted model give rise to suboptimal coding efficiency characterized by a rate-distortion relation R(D) similar in form to the theoretical rate-distortion function. Numerical examples of this performance measure are presented for Fourier, Walsh, Haar, and Karhunen-Loeve transforms.

  14. Comparison of computer codes for evaluation of double-supply-frequency pulsations in linear induction pumps

    International Nuclear Information System (INIS)

    Kirillov, Igor R.; Obukhov, Denis M.; Ogorodnikov, Anatoly P.; Araseki, Hideo

    2004-01-01

    The paper describes and compares three computer codes that are able to estimate the double-supply-frequency (DSF) pulsations in annular linear induction pumps (ALIPs). The DSF pulsations are the result of interaction of the magnetic field and induced in liquid metal currents both changing with supply-frequency. They may be of some concern for electromagnetic pumps (EMP) exploitation and need to be evaluated at their design. The results of computer simulation are compared with experimental ones for annular linear induction pump ALIP-1

  15. Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

    International Nuclear Information System (INIS)

    Basoglu, Benan; Yamamoto, Toshihiro; Okuno, Hiroshi; Nomura, Yasushi

    1998-03-01

    In this work, we report on the development of a new computer code named TRACE for predicting the excursion characteristics of criticality excursions involving fissile solutions. TRACE employs point neutronics coupled with simple thermal-hydraulics. The temperature, the radiolytic gas effects, and the boiling phenomena are estimated using the transient heat conduction equation, a lumped-parameter energy model, and a simple boiling model, respectively. To evaluate the model, we compared our results with the results of CRAC experiments. The agreement in these comparisons is quite satisfactory. (author)

  16. Development of Nuclear Energy Security Code

    International Nuclear Information System (INIS)

    Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.

    1990-01-01

    In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)

  17. Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

    2016-11-01

    Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

  18. Variable weight Khazani-Syed code using hybrid fixed-dynamic technique for optical code division multiple access system

    Science.gov (United States)

    Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah

    2016-10-01

    Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.

  19. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  20. Coding and transmission of subband coded images on the Internet

    Science.gov (United States)

    Wah, Benjamin W.; Su, Xiao

    2001-09-01

    Subband-coded images can be transmitted in the Internet using either the TCP or the UDP protocol. Delivery by TCP gives superior decoding quality but with very long delays when the network is unreliable, whereas delivery by UDP has negligible delays but with degraded quality when packets are lost. Although images are delivered currently over the Internet by TCP, we study in this paper the use of UDP to deliver multi-description reconstruction-based subband-coded images. First, in order to facilitate recovery from UDP packet losses, we propose a joint sender-receiver approach for designing optimized reconstruction-based subband transform (ORB-ST) in multi-description coding (MDC). Second, we carefully evaluate the delay-quality trade-offs between the TCP delivery of SDC images and the UDP and combined TCP/UDP delivery of MDC images. Experimental results show that our proposed ORB-ST performs well in real Internet tests, and UDP and combined TCP/UDP delivery of MDC images provide a range of attractive alternatives to TCP delivery.

  1. Dichorionic twin ultrasound surveillance: sonography every 4 weeks significantly underperforms sonography every 2 weeks: results of the Prospective Multicenter ESPRiT Study.

    Science.gov (United States)

    Corcoran, Siobhan; Breathnach, Fionnuala; Burke, Gerard; McAuliffe, Fionnuala; Geary, Michael; Daly, Sean; Higgins, John; Hunter, Alyson; Morrison, John J; Higgins, Shane; Mahony, Rhona; Dicker, Patrick; Tully, Elizabeth; Malone, Fergal D

    2015-10-01

    A 2-week ultrasound scanning schedule for monochorionic twins is endorsed widely. There is a lack of robust data to inform a schedule for the surveillance of dichorionic gestations. We aimed to determine how ultrasound scanning that is performed at 2- or 4-week intervals (or every 4 weeks before 32 weeks' gestation and every 2 weeks thereafter) may impact the prenatal detection of fetal growth restriction (FGR) and ultimately influence timing of delivery. In a consecutive cohort of 789 dichorionic twin pregnancies that were recruited prospectively for the multicenter Evaluation of Sonographic Predictors of Restricted Growth in Twins study, ultrasound determination of fetal growth and interrogation of umbilical and middle cerebral artery Doppler scans were performed every 2 weeks from 24 weeks' gestation until delivery. Complete delivery and perinatal outcome data were recorded for all pregnancies. Where delivery was prompted by FGR, abnormal umbilical artery Doppler examination or poor biophysical profile and in the absence of ruptured membranes, onset of labor, preeclampsia, or antepartum hemorrhage, the delivery was considered "ultrasound-indicated." For ultrasound-indicated deliveries, detection probabilities for FGR/abnormal umbilical artery Doppler scans/poor biophysical were determined according to the interval between examinations, by the suppression if alternate examination data. Among 789 dichorionic twin pregnancies, 66 pairs (8%) had an "ultrasound indicated" delivery. Detection of FGR was reduced from 88-69%, and detection of abnormal umbilical artery Doppler was reduced from 82-62% when a 4-week ultrasound schedule was simulated. Both of these reductions reached statistical significance. There was a nonsignificant trend toward a reduction in the recording of oligohydramnios with a 4-week interval between examinations. This study suggests that the ultrasound surveillance program of every 2 weeks that is recommended currently for monochorionic twins

  2. Placental determinants of fetal growth: identification of key factors in the insulin-like growth factor and cytokine systems using artificial neural networks

    Directory of Open Access Journals (Sweden)

    Faleschini Elena

    2008-06-01

    Full Text Available Abstract Background Changes and relationships of components of the cytokine and IGF systems have been shown in placenta and cord serum of fetal growth restricted (FGR compared with normal newborns (AGA. This study aimed to analyse a data set of clinical and biochemical data in FGR and AGA newborns to assess if a mathematical model existed and was capable of identifying these two different conditions in order to identify the variables which had a mathematically consistent biological relevance to fetal growth. Methods Whole villous tissue was collected at birth from FGR (N = 20 and AGA neonates (N = 28. Total RNA was extracted, reverse transcribed and then real-time quantitative (TaqMan RT-PCR was performed to quantify cDNA for IGF-I, IGF-II, IGFBP-1, IGFBP-2 and IL-6. The corresponding proteins with TNF-α in addition were assayed in placental lysates using specific kits. The data were analysed using Artificial Neural Networks (supervised networks, and principal component analysis and connectivity map. Results The IGF system and IL-6 allowed to predict FGR in approximately 92% of the cases and AGA in 85% of the cases with a low number of errors. IGF-II, IGFBP-2, and IL-6 content in the placental lysates were the most important factors connected with FGR. The condition of being FGR was connected mainly with the IGF-II placental content, and the latter with IL-6 and IGFBP-2 concentrations in placental lysates. Conclusion These results suggest that further research in humans should focus on these biochemical data. Furthermore, this study offered a critical revision of previous studies. The understanding of this system biology is relevant to the development of future therapeutical interventions possibly aiming at reducing IL-6 and IGFBP-2 concentrations preserving IGF bioactivity in both placenta and fetus.

  3. Reaction behavior of SO2 in the sintering process with flue gas recirculation.

    Science.gov (United States)

    Yu, Zhi-Yuan; Fan, Xiao-Hui; Gan, Min; Chen, Xu-Ling; Chen, Qiang; Huang, Yun-Song

    2016-07-01

    The primary goal of this paper is to reveal the reaction behavior of SO2 in the sinter zone, combustion zone, drying-preheating zone, and over-wet zone during flue gas recirculation (FGR) technique. The results showed that SO2 retention in the sinter zone was associated with free-CaO in the form of CaSO3/CaSO4, and the SO2 adsorption reached a maximum under 900ºC. SO2 in the flue gas came almost from the combustion zone. One reaction behavior was the oxidation of sulfur in the sintering mix when the temperature was between 800 and 1000ºC; the other behavior was the decomposition of sulfite/sulfate when the temperature was over 1000ºC. However, the SO2 adsorption in the sintering bed mainly occurred in the drying-preheating zone, adsorbed by CaCO3, Ca(OH)2, and CaO. When the SO2 adsorption reaction in the drying-preheating zone reached equilibrium, the excess SO2 gas continued to migrate to the over-wet zone and was then absorbed by Ca(OH)2 and H2O. The emission rising point of SO2 moved forward in combustion zone, and the concentration of SO2 emissions significantly increased in the case of flue gas recirculation (FGR) technique. Aiming for the reuse of the sensible heat and a reduction in exhaust gas emission, the FGR technique is proposed in the iron ore sintering process. When using the FGR technique, SO2 emission in exhaust gas gets changed. In practice, the application of the FGR technique in a sinter plant should be cooperative with the flue gas desulfurization (FGD) technique. Thus, it is necessary to study the influence of the FGR technique on SO2 emissions because it will directly influence the demand and design of the FGD system.

  4. LDPC concatenated space-time block coded system in multipath fading environment: Analysis and evaluation

    Directory of Open Access Journals (Sweden)

    Surbhi Sharma

    2011-06-01

    Full Text Available Irregular low-density parity-check (LDPC codes have been found to show exceptionally good performance for single antenna systems over a wide class of channels. In this paper, the performance of LDPC codes with multiple antenna systems is investigated in flat Rayleigh and Rician fading channels for different modulation schemes. The focus of attention is mainly on the concatenation of irregular LDPC codes with complex orthogonal space-time codes. Iterative decoding is carried out with a density evolution method that sets a threshold above which the code performs well. For the proposed concatenated system, the simulation results show that the QAM technique achieves a higher coding gain of 8.8 dB and 3.2 dB over the QPSK technique in Rician (LOS and Rayleigh (NLOS faded environments respectively.

  5. Status of the ASTEC integral code

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Jacq, F.; Allelein, H.J.

    2000-01-01

    The ASTEC (Accident Source Term Evaluation Code) integrated code is developed since 1997 in close collaboration by IPSN and GRS to predict an entire LWR severe accident sequence from the initiating event up to Fission Product (FP) release out of the containment. The applications of such a code are source term determination studies, scenario evaluations, accident management studies and Probabilistic Safety Assessment level 2 (PSA-2) studies. The version V0 of ASTEC is based on the RCS modules of the ESCADRE integrated code (IPSN) and on the upgraded RALOC and FIPLOC codes (GRS) for containment thermalhydraulics and aerosol behaviour. The latest version V0.2 includes the general feed-back from the overall validation performed in 1998 (25 separate-effect experiments, PHEBUS.FP FPT1 integrated experiment), some modelling improvements (i.e. silver-iodine reactions in the containment sump), and the implementation of the main safety systems for Severe Accident Management. Several reactor-applications are under way on French and German PWR, and on VVER-1000, all with a multi-compartment configuration of the containment. The total IPSN-GRS manpower involved in ASTEC project is today about 20 men/year. The main evolution of the next version V1, foreseen end of 2001, concerns the integration of the front-end phase and the improvement of the in-vessel degradation late-phase modelling. (author)

  6. Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

    2016-11-15

    Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

  7. SEAPATH: A microcomputer code for evaluating physical security effectiveness using adversary sequence diagrams

    International Nuclear Information System (INIS)

    Darby, J.L.

    1986-01-01

    The Adversary Sequence Diagram (ASD) concept was developed by Sandia National Laboratories (SNL) to examine physical security system effectiveness. Sandia also developed a mainframe computer code, PANL, to analyze the ASD. The authors have developed a microcomputer code, SEAPATH, which also analyzes ASD's. The Authors are supporting SNL in software development of the SAVI code; SAVI utilizes the SEAPATH algorithm to identify and quantify paths

  8. Dosimetric Significance of the ICRP's Updated Guidance and Models, 1989-2003, and Implications for U.S. Federal Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Leggett, R.W.

    2003-09-10

    Over the past two decades the U.S. Environmental Protection Agency (EPA) has issued a series of Federal guidance documents for the purpose of providing the Federal and State agencies with technical information to assist their implementation of radiation protection programs. Currently recommended dose conversion factors, annual limits on intake, and derived air concentrations for intake of radionuclides are tabulated in Federal Guidance Report No. 11 (FGR 11), published in 1988. The tabulations in FGR 11 were based on dosimetric quantities and biokinetic and dosimetric models of the International Commission on Radiological Protection (ICRP) developed for application to occupational exposures. Since the publication of FGR 11 the ICRP has revised some of its dosimetric quantities and its models for workers and has also developed age-specific models and dose conversion factors for intake of radionuclides by members of the public. This report examines the extent of the changes in the inhalation and ingestion dose coefficients of FGR 11 implied by the updated recommendations of the ICRP, both for workers and members of the public.

  9. Diagnostic accuracy of fundal height and handheld ultrasound-measured abdominal circumference to screen for fetal growth abnormalities

    Science.gov (United States)

    Haragan, Adriane F.; Hulsey, Thomas C.; Hawk, Angela F.; Newman, Roger B.; Chang, Eugene Y.

    2015-01-01

    OBJECTIVE We sought to compare fundal height and handheld ultrasound–measured fetal abdominal circumference (HHAC) for the prediction of fetal growth restriction (FGR) or large for gestational age. STUDY DESIGN This was a diagnostic accuracy study in nonanomalous singleton pregnancies between 24 and 40 weeks’ gestation. Patients underwent HHAC and fundal height measurement prior to formal growth ultrasound. FGR was defined as estimated fetal weight less than 10%, whereas large for gestational age was defined as estimated fetal weight greater than 90%. Sensitivity and specificity were calculated and compared using methods described elsewhere. RESULTS There were 251 patients included in this study. HHAC had superior sensitivity and specificity for the detection of FGR (sensitivity, 100% vs 42.86%) and (specificity, 92.62% vs 85.24%). HHAC had higher specificity but lower sensitivity when screening for LGA (specificity, 85.66% vs 66.39%) and (sensitivity, 57.14% vs 71.43%). CONCLUSION HHAC could prove to be a valuable screening tool in the detection of FGR. Further studies are needed in a larger population. PMID:25818672

  10. Role of the placental Vitamin D receptor in modulating feto-placental growth in Fetal growth restriction and Preeclampsia-affected pregnancies.

    Directory of Open Access Journals (Sweden)

    Padma eMurthi

    2016-02-01

    Full Text Available Fetal growth restriction (FGR is a common pregnancy complication that affects up to 5% of pregnancies worldwide. Recent studies demonstrate that Vitamin D deficiency is implicated in reduced fetal growth, which may be rescued by supplementation of Vitamin D. Despite this, the pathway(s by which Vitamin D modulate fetal growth remains to be investigated. Our own studies demonstrate that the Vitamin D receptor (VDR is significantly decreased in placentae from human pregnancies complicated by FGR and contributes to abnormal placental trophoblast apoptosis and differentiation and regulation of cell-cycle genes in vitro. Thus, Vitamin D signalling is important for normal placental function and fetal growth. This review discusses the association of Vitamin D with fetal growth, the function of Vitamin D and its receptor in pregnancy, as well as the functional significance of a placental source of Vitamin D in FGR. Additionally, we propose that for Vitamin D to be clinically effective to prevent and manage FGR, the molecular mechanisms of Vitamin D and its receptor in modulating fetal growth requires further investigation.

  11. Prediction of Fetal Growth Restriction by Analyzing the Messenger RNAs of Angiogenic Factor in the Plasma of Pregnant Women.

    Science.gov (United States)

    Takenaka, Shin; Ventura, Walter; Sterrantino, Anna Freni; Kawashima, Akihiro; Koide, Keiko; Hori, Kyoko; Farina, Antonio; Sekizawa, Akihiko

    2015-06-01

    To predict the occurrence of fetal growth restriction (FGR) by analyzing messenger RNA (mRNA) expression levels of vascular endothelial growth factor receptor 1 (fms-like tyrosine kinase 1 [Flt-1]) in maternal blood. Eleven women with FGR were matched with 88 controls. Plasma samples were obtained during each trimester. The Flt-1 mRNA expression levels were compared between groups. Predicted probabilities were calculated, and sensitivity-specificity (receiver-operating characteristic [ROC]) curves were assessed based on regression models for each trimester measurement and possible combinations of measurements. The mRNA levels of the FGR group during all trimesters were significantly higher than those of the control group. The ROC curve of combined first and second trimester data yielded a detection rate of 60% at a 10% false-positive rate, with an area under curve of 0.79. The Flt-1 mRNA expression in maternal blood can be used as a marker to predict the development of FGR, long before a clinical diagnosis is made. © The Author(s) 2014.

  12. Ethical Code Effectiveness in Football Clubs: A Longitudinal Analysis

    OpenAIRE

    Constandt, Bram; De Waegeneer, Els; Willem, Annick

    2017-01-01

    As football (soccer) clubs are facing different ethical challenges, many clubs are turning to ethical codes to counteract unethical behaviour. However, both in- and outside the sport field, uncertainty remains about the effectiveness of these ethical codes. For the first time, a longitudinal study design was adopted to evaluate code effectiveness. Specifically, a sample of non-professional football clubs formed the subject of our inquiry. Ethical code effectiveness was...

  13. NRC model simulations in support of the hydrologic code intercomparison study (HYDROCOIN): Level 1-code verification

    International Nuclear Information System (INIS)

    1988-03-01

    HYDROCOIN is an international study for examining ground-water flow modeling strategies and their influence on safety assessments of geologic repositories for nuclear waste. This report summarizes only the combined NRC project temas' simulation efforts on the computer code bench-marking problems. The codes used to simulate thesee seven problems were SWIFT II, FEMWATER, UNSAT2M USGS-3D, AND TOUGH. In general, linear problems involving scalars such as hydraulic head were accurately simulated by both finite-difference and finite-element solution algorithms. Both types of codes produced accurate results even for complex geometrics such as intersecting fractures. Difficulties were encountered in solving problems that invovled nonlinear effects such as density-driven flow and unsaturated flow. In order to fully evaluate the accuracy of these codes, post-processing of results using paricle tracking algorithms and calculating fluxes were examined. This proved very valuable by uncovering disagreements among code results even through the hydraulic-head solutions had been in agreement. 9 refs., 111 figs., 6 tabs

  14. SCRAM reactivity calculations with the KIKO3D code

    International Nuclear Information System (INIS)

    Hordosy, G.; Kerszturi, A.; Maraczy, Cs.; Temesvari, E.

    1999-01-01

    Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code, The time and space dependent neutron flux in the reactor core during the rod drop measurement was calculated by the KIKO3D nodal diffusion code. For calculating the ionisation chamber signals the Green function technique was applied. The Green functions of ionisation chambers were evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals during asymmetric SCRAM measurements were calculated and compared with measured data using the inverse point kinetics transformation. The sufficient agreement validates the KIKO3D code to determine the reactivities after SCRAM. (Authors)

  15. SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 1, Part 2: Control modules S1--H1; Revision 5

    International Nuclear Information System (INIS)

    1997-03-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system

  16. SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 2, Part 3: Functional modules F16--F17; Revision 5

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system.

  17. SCALE: A modular code system for performing Standardized Computer Analyses for Licensing Evaluation. Volume 2, Part 3: Functional modules F16--F17; Revision 5

    International Nuclear Information System (INIS)

    1997-03-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automated the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.3 of the system

  18. Review and evaluation of technology, equipment, codes and standards for digitization of industrial radiographic film

    International Nuclear Information System (INIS)

    1992-05-01

    This reports contains a review and evaluation of the technology, equipment, and codes and standards related to the digitization of industrial radiographic film. The report presents recommendations and equipment-performance specifications that will allow the digitization of radiographic film from nuclear power plant components in order to produce faithful reproductions of flaw images of interest on the films. Justification for the specifications selected are provided. Performance demonstration tests for the digitization process are required and criteria for such tests is presented. Also several comments related to implementation of the technology are presented and discussed

  19. Comparative evaluation of various optimization methods and the development of an optimization code system SCOOP

    International Nuclear Information System (INIS)

    Suzuki, Tadakazu

    1979-11-01

    Thirty two programs for linear and nonlinear optimization problems with or without constraints have been developed or incorporated, and their stability, convergence and efficiency have been examined. On the basis of these evaluations, the first version of the optimization code system SCOOP-I has been completed. The SCOOP-I is designed to be an efficient, reliable, useful and also flexible system for general applications. The system enables one to find global optimization point for a wide class of problems by selecting the most appropriate optimization method built in it. (author)

  20. CATHARE code development and assessment methodologies

    International Nuclear Information System (INIS)

    Micaelli, J.C.; Barre, F.; Bestion, D.

    1995-01-01

    The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

  1. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  2. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  3. Nuclear data libraries for Tripoli-3.5 code; Bibliotheques de donnees nucleaires pour le code tripoli-3.5

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th

    2001-07-01

    The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10{sup -5} eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

  4. Evaluation of the efficacy of twelve mitochondrial protein-coding genes as barcodes for mollusk DNA barcoding.

    Science.gov (United States)

    Yu, Hong; Kong, Lingfeng; Li, Qi

    2016-01-01

    In this study, we evaluated the efficacy of 12 mitochondrial protein-coding genes from 238 mitochondrial genomes of 140 molluscan species as potential DNA barcodes for mollusks. Three barcoding methods (distance, monophyly and character-based methods) were used in species identification. The species recovery rates based on genetic distances for the 12 genes ranged from 70.83 to 83.33%. There were no significant differences in intra- or interspecific variability among the 12 genes. The monophyly and character-based methods provided higher resolution than the distance-based method in species delimitation. Especially in closely related taxa, the character-based method showed some advantages. The results suggested that besides the standard COI barcode, other 11 mitochondrial protein-coding genes could also be potentially used as a molecular diagnostic for molluscan species discrimination. Our results also showed that the combination of mitochondrial genes did not enhance the efficacy for species identification and a single mitochondrial gene would be fully competent.

  5. Preclinical evaluation of spatial frequency domain-enabled wide-field quantitative imaging for enhanced glioma resection

    Science.gov (United States)

    Sibai, Mira; Fisher, Carl; Veilleux, Israel; Elliott, Jonathan T.; Leblond, Frederic; Roberts, David W.; Wilson, Brian C.

    2017-07-01

    5-Aminolevelunic acid-induced protoporphyrin IX (PpIX) fluorescence-guided resection (FGR) enables maximum safe resection of glioma by providing real-time tumor contrast. However, the subjective visual assessment and the variable intrinsic optical attenuation of tissue limit this technique to reliably delineating only high-grade tumors that display strong fluorescence. We have previously shown, using a fiber-optic probe, that quantitative assessment using noninvasive point spectroscopic measurements of the absolute PpIX concentration in tissue further improves the accuracy of FGR, extending it to surgically curable low-grade glioma. More recently, we have shown that implementing spatial frequency domain imaging with a fluorescent-light transport model enables recovery of two-dimensional images of [PpIX], alleviating the need for time-consuming point sampling of the brain surface. We present first results of this technique modified for in vivo imaging on an RG2 rat brain tumor model. Despite the moderate errors in retrieving the absorption and reduced scattering coefficients in the subdiffusive regime of 14% and 19%, respectively, the recovered [PpIX] maps agree within 10% of the point [PpIX] values measured by the fiber-optic probe, validating its potential as an extension or an alternative to point sampling during glioma resection.

  6. System Level Evaluation of Innovative Coded MIMO-OFDM Systems for Broadcasting Digital TV

    Directory of Open Access Journals (Sweden)

    Y. Nasser

    2008-01-01

    Full Text Available Single-frequency networks (SFNs for broadcasting digital TV is a topic of theoretical and practical interest for future broadcasting systems. Although progress has been made in the characterization of its description, there are still considerable gaps in its deployment with MIMO technique. The contribution of this paper is multifold. First, we investigate the possibility of applying a space-time (ST encoder between the antennas of two sites in SFN. Then, we introduce a 3D space-time-space block code for future terrestrial digital TV in SFN architecture. The proposed 3D code is based on a double-layer structure designed for intercell and intracell space time-coded transmissions. Eventually, we propose to adapt a technique called effective exponential signal-to-noise ratio (SNR mapping (EESM to predict the bit error rate (BER at the output of the channel decoder in the MIMO systems. The EESM technique as well as the simulations results will be used to doubly check the efficiency of our 3D code. This efficiency is obtained for equal and unequal received powers whatever is the location of the receiver by adequately combining ST codes. The 3D code is then a very promising candidate for SFN architecture with MIMO transmission.

  7. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

  8. 21 CFR 201.25 - Bar code label requirements.

    Science.gov (United States)

    2010-04-01

    ... Number/Uniform Code Council (EAN.UCC) or Health Industry Business Communications Council (HIBCC... alternative regulatory program or method of product use renders the bar code unnecessary for patient safety... Evaluation and Research, Food and Drug Administration, 5600 Fishers Lane, Rockville, MD 20857 (requests...

  9. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

    2014-01-01

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems

  10. The analysis of thermal-hydraulic models in MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)

    1996-07-15

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.

  11. Assessing the INTERTRAN code for application in Asian environs

    International Nuclear Information System (INIS)

    Yoshimura, S.

    1986-10-01

    A Japanese study, which was carried out as part of the IAEA Coordinated Research Programme on Radiation Protection Implications of Transport Accidents Involving Radioactive Materials, provided evaluations of transport conditions of nuclear fuel in Japan. Nuclear fuel is transported in Japan in the form of UO 2 , UF 6 , fresh fuel assemblies and spent fuel. Based on these transport conditions calculations were made using the INTERTRAN code which was developed as part of the IAEA Coordinated Research Programme on Safe Transport of Radioactive Materials (1980-1985), for assessing doses to workers and to the public due to the transport of nuclear fuel. As a part of the study, a new code was developed for evaluating radiological impacts of the transport of radioactive materials. The code was also used for assessing doses from the transport of nuclear fuel in Japan. The results indicate that doses to workers and to the public due to the incident-free transport of nuclear fuel are low, i.e., of the order of 1-30 man mSv/100 km. The doses calculated by the Japanese code were in general slightly smaller than the doses calculated using the INTERTRAN code. The study concerned normal conditions of transport, i.e., no impact from incidents or accidents was evaluated. The study resulted, in addition, in some suggestions for further developing the INTERTRAN code

  12. An Optimal Linear Coding for Index Coding Problem

    OpenAIRE

    Pezeshkpour, Pouya

    2015-01-01

    An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

  13. Evaluating the Coding and Workload Accounting Improvement Initiative of Madigan Army Medical Center

    National Research Council Canada - National Science Library

    Bewley, Lee W; Bender, Brian J

    2007-01-01

    ... documentation, provider coding accuracy and education, and clinic electronic medical record (AHLTA) usage. The desired end state of the CWAI is improved medical documentation and coding accuracy at MAMC...

  14. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  15. GARLIC - A general purpose atmospheric radiative transfer line-by-line infrared-microwave code: Implementation and evaluation

    Science.gov (United States)

    Schreier, Franz; Gimeno García, Sebastián; Hedelt, Pascal; Hess, Michael; Mendrok, Jana; Vasquez, Mayte; Xu, Jian

    2014-04-01

    A suite of programs for high resolution infrared-microwave atmospheric radiative transfer modeling has been developed with emphasis on efficient and reliable numerical algorithms and a modular approach appropriate for simulation and/or retrieval in a variety of applications. The Generic Atmospheric Radiation Line-by-line Infrared Code - GARLIC - is suitable for arbitrary observation geometry, instrumental field-of-view, and line shape. The core of GARLIC's subroutines constitutes the basis of forward models used to implement inversion codes to retrieve atmospheric state parameters from limb and nadir sounding instruments. This paper briefly introduces the physical and mathematical basics of GARLIC and its descendants and continues with an in-depth presentation of various implementation aspects: An optimized Voigt function algorithm combined with a two-grid approach is used to accelerate the line-by-line modeling of molecular cross sections; various quadrature methods are implemented to evaluate the Schwarzschild and Beer integrals; and Jacobians, i.e. derivatives with respect to the unknowns of the atmospheric inverse problem, are implemented by means of automatic differentiation. For an assessment of GARLIC's performance, a comparison of the quadrature methods for solution of the path integral is provided. Verification and validation are demonstrated using intercomparisons with other line-by-line codes and comparisons of synthetic spectra with spectra observed on Earth and from Venus.

  16. Differential correlations between maternal hair levels of tobacco and alcohol with fetal growth restriction clinical subtypes.

    Science.gov (United States)

    Sabra, Sally; Malmqvist, Ebba; Almeida, Laura; Gratacos, Eduard; Gomez Roig, Maria Dolores

    2018-08-01

    Maternal exposure to tobacco and alcohol is a known cause, among others, for fetal growth restriction (FGR). Clinically, FGR can be subclassified into two forms: intrauterine growth restriction (IUGR) and small for gestational age (SGA), based on the severity of the growth retardation, and abnormal uterine artery Doppler or cerebro-placental ratio. This study aimed at investigating any differential correlation between maternal exposures to these toxins with the two clinical forms of FGR. Therefore, a case-control study was conducted in Barcelona, Spain. Sixty-four FGR subjects, who were further subclassified into IUGR (n = 36) and SGA (n = 28), and 89 subjects matched appropriate-for-gestational age (AGA), were included. The levels of nicotine (NIC) and ethyl glucuronide (EtG), biomarkers of tobacco and alcohol exposure, respectively, were assessed in the maternal hair in the third trimester. Our analysis showed 65% of the pregnant women consumed alcohol, 25% smoked, and 19% did both. The odds ratios (ORs) of IUGR were 21 times versus 14 times for being SGA with maternal heavy smoking, while with alcohol consumption the ORs for IUGR were 22 times versus 37 times for the SGA group. The differential correlations between these toxins with the two subtypes of FGR suggest different mechanisms influencing fetal weight. Our alarming data of alcohol consumption during pregnancy should be considered for further confirmation among Spanish women. Copyright © 2018 Elsevier Inc. All rights reserved.

  17. User Instructions for the CiderF Individual Dose Code and Associated Utility Codes

    Energy Technology Data Exchange (ETDEWEB)

    Eslinger, Paul W.; Napier, Bruce A.

    2013-08-30

    Historical activities at facilities producing nuclear materials for weapons released radioactivity into the air and water. Past studies in the United States have evaluated the release, atmospheric transport and environmental accumulation of 131I from the nuclear facilities at Hanford in Washington State and the resulting dose to members of the public (Farris et al. 1994). A multi-year dose reconstruction effort (Mokrov et al. 2004) is also being conducted to produce representative dose estimates for members of the public living near Mayak, Russia, from atmospheric releases of 131I at the facilities of the Mayak Production Association. The approach to calculating individual doses to members of the public from historical releases of airborne 131I has the following general steps: • Construct estimates of releases 131I to the air from production facilities. • Model the transport of 131I in the air and subsequent deposition on the ground and vegetation. • Model the accumulation of 131I in soil, water and food products (environmental media). • Calculate the dose for an individual by matching the appropriate lifestyle and consumption data for the individual to the concentrations of 131I in environmental media at their residence location. A number of computer codes were developed to facilitate the study of airborne 131I emissions at Hanford. The RATCHET code modeled movement of 131I in the atmosphere (Ramsdell Jr. et al. 1994). The DECARTES code modeled accumulation of 131I in environmental media (Miley et al. 1994). The CIDER computer code estimated annual doses to individuals (Eslinger et al. 1994) using the equations and parameters specific to Hanford (Snyder et al. 1994). Several of the computer codes developed to model 131I releases from Hanford are general enough to be used for other facilities. This document provides user instructions for computer codes calculating doses to members of the public from atmospheric 131I that have two major differences from the

  18. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  19. Appraisal of the PREP, KITT, and SAMPLE computer codes for the evaluation of the reliability characteristics of engineered systems

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, P; White, R F

    1976-01-01

    For the probabilistic approach to reactor safety assessment by the use of event tree and fault tree techniques it is essential to be able to estimate the probabilities of failure of the various engineered safety features provided to mitigate the effects of postulated accident sequences. The PREP, KITT and SAMPLE computer codes, which incorporate Kinetic Tree Theory, perform these calculations and have been used extensively to evaluate the reliability characteristics of engineered safety features of American nuclear reactors. Working versions of these computer codes are now available in SRD, and this report explains the merits, capabilities and ease of application of the PREP, KITT, and SAMPLE programs for the solution of system reliability problems.

  20. Overview of Grid Codes for Photovoltaic Integration

    DEFF Research Database (Denmark)

    Zheng, Qianwei; Li, Jiaming; Ai, Xiaomeng

    2017-01-01

    The increasing grid-connected photovoltaic (PV) power stations might threaten the safety and stability of power system. Therefore, the grid code is developed for PV power stations to ensure the security of PV integrated power systems. In this paper, requirements for PV power integration in differ...... in different grid codes are first investigated. On this basis, the future advocacy is concluded. Finally, several evaluation indices are proposed to quantify the grid code compliance so that the system operators can validate all these requirements by simulation....

  1. Integrating Renewable Energy Requirements Into Building Energy Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kaufmann, John R.; Hand, James R.; Halverson, Mark A.

    2011-07-01

    This report evaluates how and when to best integrate renewable energy requirements into building energy codes. The basic goals were to: (1) provide a rough guide of where we’re going and how to get there; (2) identify key issues that need to be considered, including a discussion of various options with pros and cons, to help inform code deliberations; and (3) to help foster alignment among energy code-development organizations. The authors researched current approaches nationally and internationally, conducted a survey of key stakeholders to solicit input on various approaches, and evaluated the key issues related to integration of renewable energy requirements and various options to address those issues. The report concludes with recommendations and a plan to engage stakeholders. This report does not evaluate whether the use of renewable energy should be required on buildings; that question involves a political decision that is beyond the scope of this report.

  2. Fixed capacity and variable member grouping assignment of orthogonal variable spreading factor code tree for code division multiple access networks

    Directory of Open Access Journals (Sweden)

    Vipin Balyan

    2014-08-01

    Full Text Available Orthogonal variable spreading factor codes are used in the downlink to maintain the orthogonality between different channels and are used to handle new calls arriving in the system. A period of operation leads to fragmentation of vacant codes. This leads to code blocking problem. The assignment scheme proposed in this paper is not affected by fragmentation, as the fragmentation is generated by the scheme itself. In this scheme, the code tree is divided into groups whose capacity is fixed and numbers of members (codes are variable. A group with maximum number of busy members is used for assignment, this leads to fragmentation of busy groups around code tree and compactness within group. The proposed scheme is well evaluated and compared with other schemes using parameters like code blocking probability and call establishment delay. Through simulations it has been demonstrated that the proposed scheme not only adequately reduces code blocking probability, but also requires significantly less time before assignment to locate a vacant code for assignment, which makes it suitable for the real-time calls.

  3. On decoding of multi-level MPSK modulation codes

    Science.gov (United States)

    Lin, Shu; Gupta, Alok Kumar

    1990-01-01

    The decoding problem of multi-level block modulation codes is investigated. The hardware design of soft-decision Viterbi decoder for some short length 8-PSK block modulation codes is presented. An effective way to reduce the hardware complexity of the decoder by reducing the branch metric and path metric, using a non-uniform floating-point to integer mapping scheme, is proposed and discussed. The simulation results of the design are presented. The multi-stage decoding (MSD) of multi-level modulation codes is also investigated. The cases of soft-decision and hard-decision MSD are considered and their performance are evaluated for several codes of different lengths and different minimum squared Euclidean distances. It is shown that the soft-decision MSD reduces the decoding complexity drastically and it is suboptimum. The hard-decision MSD further simplifies the decoding while still maintaining a reasonable coding gain over the uncoded system, if the component codes are chosen properly. Finally, some basic 3-level 8-PSK modulation codes using BCH codes as component codes are constructed and their coding gains are found for hard decision multistage decoding.

  4. Calculation code evaluating the confinement of a nuclear facility in case of fires

    International Nuclear Information System (INIS)

    Laborde, J.C.; Prevost, C.; Vendel, J.

    1995-01-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation

  5. Calculation code evaluating the confinement of a nuclear facility in case of fires

    Energy Technology Data Exchange (ETDEWEB)

    Laborde, J.C.; Prevost, C.; Vendel, J. [and others

    1995-02-01

    Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.

  6. Validation of comprehensive space radiation transport code

    International Nuclear Information System (INIS)

    Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.

    1998-01-01

    The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation

  7. Transport code and nuclear data in intermediate energy region

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-11-01

    We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

  8. Transport code and nuclear data in intermediate energy region

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-01-01

    We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

  9. Development of the next generation code system as an engineering modeling language (6). Development of a cross section adjustment and nuclear design accuracy evaluation solver

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2008-01-01

    A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. Further, several new functions such as searching and editing calculation data are provided in the new solver. These functions can be easily customised by users because they are designed to work cooperatively with Python scripting language, which is used as a user interface of the MARBLE system. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. In addition, it is possible to reuse existing input data files prepared for the conventional code system because the new solver utilities support the conventional formats. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation

  10. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  11. User Effect on Code Application and Qualification Needs

    International Nuclear Information System (INIS)

    D'Auria, F.; Salah, A.B.

    2008-01-01

    Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

  12. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  13. Speed up of MCACE, a Monte Carlo code for evaluation of shielding safety, by parallel computer, (3)

    International Nuclear Information System (INIS)

    Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka; Onodera, Emi; Imawaka, Tsuneyuki; Yoda, Yoshihisa.

    1993-07-01

    The parallel computing of the MCACE code has been studied on two platforms; 1) Shared Memory Type Vector-Parallel Computer Monte-4 and 2) Networked Several Workstations. On the Monte-4, a disk-file has been allocated to collect all results computed by 4 CPUs in parallel, executing the copy of the MCACE code on each CPU. On the workstations under network environment, two parallel models have been evaluated; 1) a host-node model and 2) the model used on the Monte-4 where no software for parallelization has been employed but only standard FORTRAN language. The measurement of computing times has showed that speed up of about 3 times has been achieved by using 4 CPUs of the Monte-4. Further, connecting 4 workstations by network, the computing speed by parallelization has achieved faster than our scalar main frame computer, FACOM M-780. (author)

  14. Bit-wise arithmetic coding for data compression

    Science.gov (United States)

    Kiely, A. B.

    1994-01-01

    This article examines the problem of compressing a uniformly quantized independent and identically distributed (IID) source. We present a new compression technique, bit-wise arithmetic coding, that assigns fixed-length codewords to the quantizer output and uses arithmetic coding to compress the codewords, treating the codeword bits as independent. We examine the performance of this method and evaluate the overhead required when used block-adaptively. Simulation results are presented for Gaussian and Laplacian sources. This new technique could be used as the entropy coder in a transform or subband coding system.

  15. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  16. Early, Incomplete, or Preclinical Autoimmune Systemic Rheumatic Diseases and Pregnancy Outcome.

    Science.gov (United States)

    Spinillo, Arsenio; Beneventi, Fausta; Locatelli, Elena; Ramoni, Vèronique; Caporali, Roberto; Alpini, Claudia; Albonico, Giulia; Cavagnoli, Chiara; Montecucco, Carlomaurizio

    2016-10-01

    To evaluate the impact of preclinical systemic autoimmune rheumatic disorders on pregnancy outcome. In this longitudinal cohort study, patients were enrolled during the first trimester of pregnancy if they reported having had connective tissue disorder symptoms, were found to be positive for circulating autoantibodies, and on clinical evaluation were judged to have a preclinical or incomplete rheumatic disorder. The incidence of fetal growth restriction (FGR), preeclampsia, and adverse pregnancy outcomes in patients with preclinical rheumatic disorders was compared with that in selected controls, after adjustment for confounders by penalized logistic regression. Odds ratios (ORs) and 95% confidence intervals (95% CIs) were calculated. Of 5,232 women screened, 150 (2.9%) were initially diagnosed as having a suspected rheumatic disorder. After a mean ± SD postpartum follow-up of 16.7 ± 5.5 months, 64 of these women (42.7%) had no clinically apparent rheumatic disease and 86 (57.3%) had persistent symptoms and positive autoantibody results, including 10 (6.7%) who developed a definitive rheumatic disease. The incidences of preeclampsia/FGR and of small for gestational age (SGA) infants were 5.1% (23 of 450) and 9.3% (42 of 450), respectively, among controls, 12.5% (8 of 640) (OR 2.7 [95% CI 1.1-6.4]) and 18.8% (12 of 64) (OR 2.2 [95% CI 1.1-4.5]), respectively, among women with no clinically apparent disease, and 16.3% (14 of 86) (OR 3.8 [95% CI 1.9-7.7]) and 18.6% (16 of 86) (OR 2.3 [95% CI 1.2-4.3]), respectively, among those with persisting symptoms at follow-up. Mean ± SD umbilical artery Doppler pulsatility indices were higher among women with no clinically apparent disease (0.95 ± 0.2) and those with persisting symptoms (0.96 ± 0.21) than in controls (0.89 ± 0.12) (P = 0.01 and P rheumatic disorders were associated with an increased risk of FGR/preeclampsia and SGA. The impact of these findings and their utility in screening

  17. Evaluation of conservatisms and environmental effects in ASME Code, Section III, Class 1 fatigue analysis

    International Nuclear Information System (INIS)

    Deardorff, A.F.; Smith, J.K.

    1994-08-01

    This report documents the results of a study regarding the conservatisms in ASME Code Section 3, Class 1 component fatigue evaluations and the effects of Light Water Reactor (LWR) water environments on fatigue margins. After review of numerous Class 1 stress reports, it is apparent that there is a substantial amount of conservatism present in many existing component fatigue evaluations. With little effort, existing evaluations could be modified to reduce the overall predicted fatigue usage. Areas of conservatism include design transients considerably more severe than those experienced during service, conservative grouping of transients, conservatisms that have been removed in later editions of Section 3, bounding heat transfer and stress analysis, and use of the ''elastic-plastic penalty factor'' (K 3 ). Environmental effects were evaluated for two typical components that experience severe transient thermal cycling during service, based on both design transients and actual plant data. For all reasonable values of actual operating parameters, environmental effects reduced predicted margins, but fatigue usage was still bounded by the ASME Section 3 fatigue design curves. It was concluded that the potential increase in predicted fatigue usage due to environmental effects should be more than offset by decreases in predicted fatigue usage if re-analysis were conducted to reduce the conservatisms that are present in existing component fatigue evaluations

  18. Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

    International Nuclear Information System (INIS)

    Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

    1989-12-01

    The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

  19. Spike Code Flow in Cultured Neuronal Networks.

    Science.gov (United States)

    Tamura, Shinichi; Nishitani, Yoshi; Hosokawa, Chie; Miyoshi, Tomomitsu; Sawai, Hajime; Kamimura, Takuya; Yagi, Yasushi; Mizuno-Matsumoto, Yuko; Chen, Yen-Wei

    2016-01-01

    We observed spike trains produced by one-shot electrical stimulation with 8 × 8 multielectrodes in cultured neuronal networks. Each electrode accepted spikes from several neurons. We extracted the short codes from spike trains and obtained a code spectrum with a nominal time accuracy of 1%. We then constructed code flow maps as movies of the electrode array to observe the code flow of "1101" and "1011," which are typical pseudorandom sequence such as that we often encountered in a literature and our experiments. They seemed to flow from one electrode to the neighboring one and maintained their shape to some extent. To quantify the flow, we calculated the "maximum cross-correlations" among neighboring electrodes, to find the direction of maximum flow of the codes with lengths less than 8. Normalized maximum cross-correlations were almost constant irrespective of code. Furthermore, if the spike trains were shuffled in interval orders or in electrodes, they became significantly small. Thus, the analysis suggested that local codes of approximately constant shape propagated and conveyed information across the network. Hence, the codes can serve as visible and trackable marks of propagating spike waves as well as evaluating information flow in the neuronal network.

  20. Sequence Coding and Search System for licensee event reports: code listings. Volume 2

    International Nuclear Information System (INIS)

    Gallaher, R.B.; Guymon, R.H.; Mays, G.T.; Poore, W.P.; Cagle, R.J.; Harrington, K.H.; Johnson, M.P.

    1985-04-01

    Operating experience data from nuclear power plants are essential for safety and reliability analyses, especially analyses of trends and patterns. The licensee event reports (LERs) that are submitted to the Nuclear Regulatory Commission (NRC) by the nuclear power plant utilities contain much of this data. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has developed, under contract with NSIC, a system for codifying the events reported in the LERs. The primary objective of the Sequence Coding and Search System (SCSS) is to reduce the descriptive text of the LERs to coded sequences that are both computer-readable and computer-searchable. This system provides a structured format for detailed coding of component, system, and unit effects as well as personnel errors. The database contains all current LERs submitted by nuclear power plant utilities for events occurring since 1981 and is updated on a continual basis. Volume 2 contains all valid and acceptable codes used for searching and encoding the LER data. This volume contains updated material through amendment 1 to revision 1 of the working version of ORNL/NSIC-223, Vol. 2

  1. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  2. MCOR - Monte Carlo depletion code for reference LWR calculations

    International Nuclear Information System (INIS)

    Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

    2011-01-01

    Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations

  3. Evaluation of the plastic characteristics of piping products in relation to ASME code criteria

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.

    1978-07-01

    Theories and test data relevant to the plastic characteristics of piping products are presented and compared with Code Equations in NB-3652 for Class 1 piping; in NC/ND-3652.2 for Class 2 and Class 3 piping. Comparisons are made for (a) straight pipe, (b) elbows, (c) branch connections, and (d) tees. The status of data (or lack of data) for other piping components is discussed. Comparisons are made between available data and the Code equations for two typical piping materials, SA106 Grade B and SA312 TP304, for Code Design Limits, and Service Limits A, B, C, and D. Conditions under which the Code Limits cannot be shown to be conservative from available data are pointed out. Based on the results of the study, recommendations for Code revisions are presented, along with recommendations for additional work

  4. Rate-adaptive BCH codes for distributed source coding

    DEFF Research Database (Denmark)

    Salmistraro, Matteo; Larsen, Knud J.; Forchhammer, Søren

    2013-01-01

    This paper considers Bose-Chaudhuri-Hocquenghem (BCH) codes for distributed source coding. A feedback channel is employed to adapt the rate of the code during the decoding process. The focus is on codes with short block lengths for independently coding a binary source X and decoding it given its...... strategies for improving the reliability of the decoded result are analyzed, and methods for estimating the performance are proposed. In the analysis, noiseless feedback and noiseless communication are assumed. Simulation results show that rate-adaptive BCH codes achieve better performance than low...... correlated side information Y. The proposed codes have been analyzed in a high-correlation scenario, where the marginal probability of each symbol, Xi in X, given Y is highly skewed (unbalanced). Rate-adaptive BCH codes are presented and applied to distributed source coding. Adaptive and fixed checking...

  5. [Quality assurance in coding expertise of hospital cases in the German DRG system. Evaluation of inter-rater reliability in MDK expertise].

    Science.gov (United States)

    Huber, H; Brambrink, M; Funk, R; Rieger, M

    2012-10-01

    The purpose of this study was to evaluate differences in the D-DRG results of a hospital case by 2 independently coding MKD raters. Calculation of the 2-inter-rater reliability was performed by examination of the coding of individual hospital cases. The reasons for the non-agreement of the expert evaluations and suggestions to improve the process are discussed. From the expert evaluation pool of the MDK-WL a random sample of 0.7% of the 57,375 expertises was taken. Distribution equality with the basic total was tested by the χ² test or, respectively, Fisher's exact test. For the total of 402 individual hospital cases, the G-DRG case sums of 2 experts of the MDK were determined independently and the results checked for each individual case for agreement or non-agreement. The corresponding confidence intervals with standard errors were analysed to test if certain major diagnosis categories (MDC) were statistically significantly more affected by differing expertise results than others. In 280 of the total 402 tested hospital cases, the 2 MDK raters independently reached the same G-DRG results; in 122 cases the G-DRG case sums determined by the 2 raters differed (agreement 70%; CI 65.2-74.1). Different DRG results between the 2 experts occurred regularly in the entire MDC spectrum. No MDC chapter in which significant differences between the 2 raters arose could be identified. The results of our study demonstrate an almost 70% agreement in the evaluation of hospital cost accounts by 2 independently operating MDK. This result leaves room for improvement. Optimisation potentials can be recognised on the basis of the results. Potential for improvement was established in combination with regular further training and the expansion of binding internal code recommendations as well as exchange of code-relevant information among experts in internal forums. The presented model is in principle suitable for cross-border examinations within the MDK system with the advantage that

  6. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    International Nuclear Information System (INIS)

    Kim, S. B.; Kim, D. H.; Song, Y. M.

    2011-08-01

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG

  7. Self-complementary circular codes in coding theory.

    Science.gov (United States)

    Fimmel, Elena; Michel, Christian J; Starman, Martin; Strüngmann, Lutz

    2018-04-01

    Self-complementary circular codes are involved in pairing genetic processes. A maximal [Formula: see text] self-complementary circular code X of trinucleotides was identified in genes of bacteria, archaea, eukaryotes, plasmids and viruses (Michel in Life 7(20):1-16 2017, J Theor Biol 380:156-177, 2015; Arquès and Michel in J Theor Biol 182:45-58 1996). In this paper, self-complementary circular codes are investigated using the graph theory approach recently formulated in Fimmel et al. (Philos Trans R Soc A 374:20150058, 2016). A directed graph [Formula: see text] associated with any code X mirrors the properties of the code. In the present paper, we demonstrate a necessary condition for the self-complementarity of an arbitrary code X in terms of the graph theory. The same condition has been proven to be sufficient for codes which are circular and of large size [Formula: see text] trinucleotides, in particular for maximal circular codes ([Formula: see text] trinucleotides). For codes of small-size [Formula: see text] trinucleotides, some very rare counterexamples have been constructed. Furthermore, the length and the structure of the longest paths in the graphs associated with the self-complementary circular codes are investigated. It has been proven that the longest paths in such graphs determine the reading frame for the self-complementary circular codes. By applying this result, the reading frame in any arbitrary sequence of trinucleotides is retrieved after at most 15 nucleotides, i.e., 5 consecutive trinucleotides, from the circular code X identified in genes. Thus, an X motif of a length of at least 15 nucleotides in an arbitrary sequence of trinucleotides (not necessarily all of them belonging to X) uniquely defines the reading (correct) frame, an important criterion for analyzing the X motifs in genes in the future.

  8. Diagonal Eigenvalue Unity (DEU) code for spectral amplitude coding-optical code division multiple access

    Science.gov (United States)

    Ahmed, Hassan Yousif; Nisar, K. S.

    2013-08-01

    Code with ideal in-phase cross correlation (CC) and practical code length to support high number of users are required in spectral amplitude coding-optical code division multiple access (SAC-OCDMA) systems. SAC systems are getting more attractive in the field of OCDMA because of its ability to eliminate the influence of multiple access interference (MAI) and also suppress the effect of phase induced intensity noise (PIIN). In this paper, we have proposed new Diagonal Eigenvalue Unity (DEU) code families with ideal in-phase CC based on Jordan block matrix with simple algebraic ways. Four sets of DEU code families based on the code weight W and number of users N for the combination (even, even), (even, odd), (odd, odd) and (odd, even) are constructed. This combination gives DEU code more flexibility in selection of code weight and number of users. These features made this code a compelling candidate for future optical communication systems. Numerical results show that the proposed DEU system outperforms reported codes. In addition, simulation results taken from a commercial optical systems simulator, Virtual Photonic Instrument (VPI™) shown that, using point to multipoint transmission in passive optical network (PON), DEU has better performance and could support long span with high data rate.

  9. Upgrades to the WIMS-ANL code

    International Nuclear Information System (INIS)

    Woodruff, W. L.

    1998-01-01

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries

  10. Upgrades to the WIMS-ANL code

    International Nuclear Information System (INIS)

    Woodruff, W.L.; Leopando, L.S.

    1998-01-01

    The dusty old source code in WIMS-D4M has been completely rewritten to conform more closely with current FORTRAN coding practices. The revised code contains many improvements in appearance, error checking and in control of the output. The output is now tabulated to fit the typical 80 column window or terminal screen. The Segev method for resonance integral interpolation is now an option. Most of the dimension limitations have been removed and replaced with variable dimensions within a compile-time fixed container. The library is no longer restricted to the 69 energy group structure, and two new libraries have been generated for use with the code. The new libraries are both based on ENDF/B-VI data with one having the original 69 energy group structure and the second with a 172 group structure. The common source code can be used with PCs using both Windows 95 and NT, with a Linux based operating system and with UNIX based workstations. Comparisons of this version of the code to earlier evaluations with ENDF/B-V are provided, as well as, comparisons with the new libraries. (author)

  11. List Decoding of Matrix-Product Codes from nested codes: an application to Quasi-Cyclic codes

    DEFF Research Database (Denmark)

    Hernando, Fernando; Høholdt, Tom; Ruano, Diego

    2012-01-01

    A list decoding algorithm for matrix-product codes is provided when $C_1,..., C_s$ are nested linear codes and $A$ is a non-singular by columns matrix. We estimate the probability of getting more than one codeword as output when the constituent codes are Reed-Solomon codes. We extend this list...... decoding algorithm for matrix-product codes with polynomial units, which are quasi-cyclic codes. Furthermore, it allows us to consider unique decoding for matrix-product codes with polynomial units....

  12. Evaluation of MOSTAS computer code for predicting dynamic loads in two bladed wind turbines

    Science.gov (United States)

    Kaza, K. R. V.; Janetzke, D. C.; Sullivan, T. L.

    1979-01-01

    Calculated dynamic blade loads were compared with measured loads over a range of yaw stiffnesses of the DOE/NASA Mod-O wind turbine to evaluate the performance of two versions of the MOSTAS computer code. The first version uses a time-averaged coefficient approximation in conjunction with a multi-blade coordinate transformation for two bladed rotors to solve the equations of motion by standard eigenanalysis. The second version accounts for periodic coefficients while solving the equations by a time history integration. A hypothetical three-degree of freedom dynamic model was investigated. The exact equations of motion of this model were solved using the Floquet-Lipunov method. The equations with time-averaged coefficients were solved by standard eigenanalysis.

  13. Development of the vacuum system pressure responce analysis code PRAC

    International Nuclear Information System (INIS)

    Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.

    1985-03-01

    In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)

  14. European Validation of the Integral Code ASTEC (EVITA)

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Neu, K.; Dorsselaere, J.P. Van

    2005-01-01

    The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme

  15. Sending policies in dynamic wireless mesh using network coding

    DEFF Research Database (Denmark)

    Pandi, Sreekrishna; Fitzek, Frank; Pihl, Jeppe

    2015-01-01

    This paper demonstrates the quick prototyping capabilities of the Python-Kodo library for network coding based performance evaluation and investigates the problem of data redundancy in a network coded wireless mesh with opportunistic overhearing. By means of several wireless meshed architectures ...

  16. User effects on the transient system code calculations. Final report

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.

    1995-01-01

    Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them

  17. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  18. Optix: A Monte Carlo scintillation light transport code

    Energy Technology Data Exchange (ETDEWEB)

    Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

    2014-02-11

    The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

  19. Implementation of non-condensable gases condensation suppression model into the WCOBRA/TRAC-TF2 LOCA safety evaluation code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)

  20. Combinatorial neural codes from a mathematical coding theory perspective.

    Science.gov (United States)

    Curto, Carina; Itskov, Vladimir; Morrison, Katherine; Roth, Zachary; Walker, Judy L

    2013-07-01

    Shannon's seminal 1948 work gave rise to two distinct areas of research: information theory and mathematical coding theory. While information theory has had a strong influence on theoretical neuroscience, ideas from mathematical coding theory have received considerably less attention. Here we take a new look at combinatorial neural codes from a mathematical coding theory perspective, examining the error correction capabilities of familiar receptive field codes (RF codes). We find, perhaps surprisingly, that the high levels of redundancy present in these codes do not support accurate error correction, although the error-correcting performance of receptive field codes catches up to that of random comparison codes when a small tolerance to error is introduced. However, receptive field codes are good at reflecting distances between represented stimuli, while the random comparison codes are not. We suggest that a compromise in error-correcting capability may be a necessary price to pay for a neural code whose structure serves not only error correction, but must also reflect relationships between stimuli.

  1. Remote-Handled Transuranic Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2006-12-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  2. Implementation of Energy Code Controls Requirements in New Commercial Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Rosenberg, Michael I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hart, Philip R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hatten, Mike [Solarc Energy Group, LLC, Seattle, WA (United States); Jones, Dennis [Group 14 Engineering, Inc., Denver, CO (United States); Cooper, Matthew [Group 14 Engineering, Inc., Denver, CO (United States)

    2017-03-24

    Most state energy codes in the United States are based on one of two national model codes; ANSI/ASHRAE/IES 90.1 (Standard 90.1) or the International Code Council (ICC) International Energy Conservation Code (IECC). Since 2004, covering the last four cycles of Standard 90.1 updates, about 30% of all new requirements have been related to building controls. These requirements can be difficult to implement and verification is beyond the expertise of most building code officials, yet the assumption in studies that measure the savings from energy codes is that they are implemented and working correctly. The objective of the current research is to evaluate the degree to which high impact controls requirements included in commercial energy codes are properly designed, commissioned and implemented in new buildings. This study also evaluates the degree to which these control requirements are realizing their savings potential. This was done using a three-step process. The first step involved interviewing commissioning agents to get a better understanding of their activities as they relate to energy code required controls measures. The second involved field audits of a sample of commercial buildings to determine whether the code required control measures are being designed, commissioned and correctly implemented and functioning in new buildings. The third step includes compilation and analysis of the information gather during the first two steps. Information gathered during these activities could be valuable to code developers, energy planners, designers, building owners, and building officials.

  3. Amnioinfusion before 26 weeks' gestation for severe fetal growth restriction with oligohydramnios: preliminary pilot study.

    Science.gov (United States)

    Takahashi, Yuichiro; Iwagaki, Shigenori; Chiaki, Rika; Iwasa, Tomotake; Takenaka, Motoki; Kawabata, Ichiro; Itoh, Mitsuaki

    2014-03-01

    The prognosis for severe fetal growth restriction (FGR) with severe oligohydramnios before 26 weeks' gestation (WG) is currently poor; furthermore, its management is controversial. We report the innovative new management of FGR, such as therapeutic amnioinfusion and tocolysis. For FGR and severe oligohydramnios before 26 WG complicated with absent or reversed umbilical artery end-diastolic flow velocity and/or deceleration by ultrasonography, we performed transabdominal amnioinfusion with tocolysis. Cases with multiple anomalies were excluded. Survival rate and long-term prognosis were analyzed. Among 570 FGR cases, 18 were included in the study. Mean diagnosis and delivery were at 22.6 ± 2.0 and 28.7 ± 3.3 WG. Median birthweight was 625 g (-4.2 standard deviation). Final survival rate was 11/13 (85%). There were five fetal deaths. In seven cases, oligohydramnios improved. Growth was detected in 10/18 fetuses. Furthermore, 8/8 decelerations, 4/12 cases of reversed umbilical artery end-diastolic flow velocity, 7/14 cases of brain-sparing effect, and 6/13 venous Doppler abnormalities were improved. When we detected umbilical cord compression, 8/10 cases were rescued. Eleven infants were followed up for an average of 5 years; one case of cerebral palsy with normal development and 10 cases with intact motor functions without major neurological handicap were confirmed. In cases of extremely severe FGR before 26 WG with oligohydramnios and circulatory failure, amnioinfusion might be a promising, innovative tool. © 2013 The Authors. Journal of Obstetrics and Gynaecology Research © 2013 Japan Society of Obstetrics and Gynecology.

  4. Circulating cell-derived microparticles in severe preeclampsia and in fetal growth restriction.

    Science.gov (United States)

    Alijotas-Reig, Jaume; Palacio-Garcia, Carles; Farran-Codina, Immaculada; Ruiz-Romance, Mar; Llurba, Elisa; Vilardell-Tarres, Miquel

    2012-02-01

    The behavior of the circulating microparticles (cMP) in severe preeclampsia (PE) and fetal growth restriction (FGR) is disputed. METHOD OF STUDY  Non-matched case-control study. Seventy cases of severe PE/HELLP/FGR were compared to 38 healthy pregnant women. Twenty healthy non-pregnant women acted as a control. cMP were analyzed using flow cytometry. Results are given as total (annexin-A5-ANXA5+), platelet (CD41+), leukocyte (CD45+), endothelial (CD144+CD31+//CD41-), and CD41-negative cMP/μL of plasma. Antiphospholipid antibodies (aPL) were analyzed through usual methods. Platelet and endothelial cMP increased in healthy pregnant women. PE whole group (PE±FGR) showed an increase in endothelial and CD41-negative, but not in platelet-derived, cMP. Comparing PE whole group versus healthy pregnant, we found cMP levels of endothelial and CD41- had increased. The cMP results obtained in PE group were similar to those of the PE whole group. Comparing PE group to isolated FGR, significant CD41-negative cMP increase was found in PE. According to its aPL positivity, a trend to decrease in leukocyte and endothelial-derived cMP was found in PE group. Normal pregnancy is accompanied by endothelial and platelet cell activation. Endothelial cell activation has been shown in PE but not in isolated FGR. In PE, aPL may contribute to endothelial and possibly to leukocyte cell activation. © 2011 John Wiley & Sons A/S.

  5. Software Certification - Coding, Code, and Coders

    Science.gov (United States)

    Havelund, Klaus; Holzmann, Gerard J.

    2011-01-01

    We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.

  6. Discussion on LDPC Codes and Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the progress that the workgroup on Low-Density Parity-Check (LDPC) for space link coding. The workgroup is tasked with developing and recommending new error correcting codes for near-Earth, Lunar, and deep space applications. Included in the presentation is a summary of the technical progress of the workgroup. Charts that show the LDPC decoder sensitivity to symbol scaling errors are reviewed, as well as a chart showing the performance of several frame synchronizer algorithms compared to that of some good codes and LDPC decoder tests at ESTL. Also reviewed is a study on Coding, Modulation, and Link Protocol (CMLP), and the recommended codes. A design for the Pseudo-Randomizer with LDPC Decoder and CRC is also reviewed. A chart that summarizes the three proposed coding systems is also presented.

  7. Spike Code Flow in Cultured Neuronal Networks

    Directory of Open Access Journals (Sweden)

    Shinichi Tamura

    2016-01-01

    Full Text Available We observed spike trains produced by one-shot electrical stimulation with 8 × 8 multielectrodes in cultured neuronal networks. Each electrode accepted spikes from several neurons. We extracted the short codes from spike trains and obtained a code spectrum with a nominal time accuracy of 1%. We then constructed code flow maps as movies of the electrode array to observe the code flow of “1101” and “1011,” which are typical pseudorandom sequence such as that we often encountered in a literature and our experiments. They seemed to flow from one electrode to the neighboring one and maintained their shape to some extent. To quantify the flow, we calculated the “maximum cross-correlations” among neighboring electrodes, to find the direction of maximum flow of the codes with lengths less than 8. Normalized maximum cross-correlations were almost constant irrespective of code. Furthermore, if the spike trains were shuffled in interval orders or in electrodes, they became significantly small. Thus, the analysis suggested that local codes of approximately constant shape propagated and conveyed information across the network. Hence, the codes can serve as visible and trackable marks of propagating spike waves as well as evaluating information flow in the neuronal network.

  8. The structure of affective action representations: temporal binding of affective response codes.

    Science.gov (United States)

    Eder, Andreas B; Müsseler, Jochen; Hommel, Bernhard

    2012-01-01

    Two experiments examined the hypothesis that preparing an action with a specific affective connotation involves the binding of this action to an affective code reflecting this connotation. This integration into an action plan should lead to a temporary occupation of the affective code, which should impair the concurrent representation of affectively congruent events, such as the planning of another action with the same valence. This hypothesis was tested with a dual-task setup that required a speeded choice between approach- and avoidance-type lever movements after having planned and before having executed an evaluative button press. In line with the code-occupation hypothesis, slower lever movements were observed when the lever movement was affectively compatible with the prepared evaluative button press than when the two actions were affectively incompatible. Lever movements related to approach and avoidance and evaluative button presses thus seem to share a code that represents affective meaning. A model of affective action control that is based on the theory of event coding is discussed.

  9. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Petrie, L.M.; Westfall, R.M.; Bucholz, J.A.; Hermann, O.W.; Fraley, S.K. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries.

  10. SCALE: A modular code system for performing standardized computer analyses for licensing evaluation. Functional modules F1--F8 -- Volume 2, Part 1, Revision 4

    International Nuclear Information System (INIS)

    Greene, N.M.; Petrie, L.M.; Westfall, R.M.; Bucholz, J.A.; Hermann, O.W.; Fraley, S.K.

    1995-04-01

    SCALE--a modular code system for Standardized Computer Analyses Licensing Evaluation--has been developed by Oak Ridge National Laboratory at the request of the US Nuclear Regulatory Commission. The SCALE system utilizes well-established computer codes and methods within standard analysis sequences that (1) allow an input format designed for the occasional user and/or novice, (2) automate the data processing and coupling between modules, and (3) provide accurate and reliable results. System development has been directed at problem-dependent cross-section processing and analysis of criticality safety, shielding, heat transfer, and depletion/decay problems. Since the initial release of SCALE in 1980, the code system has been heavily used for evaluation of nuclear fuel facility and package designs. This revision documents Version 4.2 of the system. The manual is divided into three volumes: Volume 1--for the control module documentation; Volume 2--for functional module documentation; and Volume 3--for documentation of the data libraries and subroutine libraries

  11. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.

  12. The VULKIN code used for evaluation of the cladding tube's performance

    International Nuclear Information System (INIS)

    Marbach, G.

    1979-01-01

    Full text: 1 - Introduction. The French approach for fast subassembly project is to analyse each component part of the subassembly and each basic phenomenon to estimate the total behaviour. The VULKIN code describes the mechanical behaviour of a clad alone. A cladding damage parameter is calculated from the observed deformations. When it is greater than a fixed value we consider that the rupture probability is not negligible. But this function is not the only limit for the irradiation project. Other limits are bound to other problems: no fuel melting bundle, interaction behaviour. 2 - VULKIN code - Presentation. The VULKIN code gives the evolution of stresses and strains distribution in the thickness of the clad with the hypothesis of revolution symmetry. This program takes into account temperature dilatation and radial thermal gradient, fission gas pressure and steel swelling due to neutron flux. The fuel clad mechanical interaction is not described by this model. Experimental results show that its influence is negligible for the most unusual subassemblies but, if it is necessary, a special calculation is obtained using a specific code like TUREN, described in another paper. This model does not consider the stresses and strains resulting from interaction between bundle and wrapper. Another model describes the bundle behaviour and determines diametral deformation limit from the subassembly geometrical characteristics. The clad is considered as an elasto-plastic element. Plastic flows instantaneous, thermal creep or irradiation creep are determined at each time. The data of this code are the geometry, the irradiation parameters (temperature, dose), the fission gas pressure evolution, the swelling law and the experimental relations for thermal and irradiation creep. The mechanical resolution is classical: the clad is divided into concentric rings. At each time the equations resulting from the equilibrium of strengths and compatibility of displacements are resolved

  13. Low-Rank Sparse Coding for Image Classification

    KAUST Repository

    Zhang, Tianzhu; Ghanem, Bernard; Liu, Si; Xu, Changsheng; Ahuja, Narendra

    2013-01-01

    In this paper, we propose a low-rank sparse coding (LRSC) method that exploits local structure information among features in an image for the purpose of image-level classification. LRSC represents densely sampled SIFT descriptors, in a spatial neighborhood, collectively as low-rank, sparse linear combinations of code words. As such, it casts the feature coding problem as a low-rank matrix learning problem, which is different from previous methods that encode features independently. This LRSC has a number of attractive properties. (1) It encourages sparsity in feature codes, locality in codebook construction, and low-rankness for spatial consistency. (2) LRSC encodes local features jointly by considering their low-rank structure information, and is computationally attractive. We evaluate the LRSC by comparing its performance on a set of challenging benchmarks with that of 7 popular coding and other state-of-the-art methods. Our experiments show that by representing local features jointly, LRSC not only outperforms the state-of-the-art in classification accuracy but also improves the time complexity of methods that use a similar sparse linear representation model for feature coding.

  14. Low-Rank Sparse Coding for Image Classification

    KAUST Repository

    Zhang, Tianzhu

    2013-12-01

    In this paper, we propose a low-rank sparse coding (LRSC) method that exploits local structure information among features in an image for the purpose of image-level classification. LRSC represents densely sampled SIFT descriptors, in a spatial neighborhood, collectively as low-rank, sparse linear combinations of code words. As such, it casts the feature coding problem as a low-rank matrix learning problem, which is different from previous methods that encode features independently. This LRSC has a number of attractive properties. (1) It encourages sparsity in feature codes, locality in codebook construction, and low-rankness for spatial consistency. (2) LRSC encodes local features jointly by considering their low-rank structure information, and is computationally attractive. We evaluate the LRSC by comparing its performance on a set of challenging benchmarks with that of 7 popular coding and other state-of-the-art methods. Our experiments show that by representing local features jointly, LRSC not only outperforms the state-of-the-art in classification accuracy but also improves the time complexity of methods that use a similar sparse linear representation model for feature coding.

  15. RH-TRU Waste Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2007-07-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  16. Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes

    International Nuclear Information System (INIS)

    Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

    2013-01-01

    The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)

  17. Variable weight spectral amplitude coding for multiservice OCDMA networks

    Science.gov (United States)

    Seyedzadeh, Saleh; Rahimian, Farzad Pour; Glesk, Ivan; Kakaee, Majid H.

    2017-09-01

    The emergence of heterogeneous data traffic such as voice over IP, video streaming and online gaming have demanded networks with capability of supporting quality of service (QoS) at the physical layer with traffic prioritisation. This paper proposes a new variable-weight code based on spectral amplitude coding for optical code-division multiple-access (OCDMA) networks to support QoS differentiation. The proposed variable-weight multi-service (VW-MS) code relies on basic matrix construction. A mathematical model is developed for performance evaluation of VW-MS OCDMA networks. It is shown that the proposed code provides an optimal code length with minimum cross-correlation value when compared to other codes. Numerical results for a VW-MS OCDMA network designed for triple-play services operating at 0.622 Gb/s, 1.25 Gb/s and 2.5 Gb/s are considered.

  18. AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, M. de.

    1984-01-01

    A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

  19. Code Generation from Pragmatics Annotated Coloured Petri Nets

    DEFF Research Database (Denmark)

    Simonsen, Kent Inge

    limited work has been done on transforming CPN model to protocol implementations. The goal of the thesis is to be able to automatically generate high-quality implementations of communication protocols based on CPN models. In this thesis, we develop a methodology for generating implementations of protocols...... third party libraries and the code should be easily usable by third party code. Finally, the code should be readable by developers with expertise on the considered platforms. In this thesis, we show that our code generation approach is able to generate code for a wide range of platforms without altering...... such as games and rich web applications. Finally, we conclude the evaluation of the criteria of our approach by using the WebSocket PA-CPN model to show that we are able to verify fairly large protocols....

  20. The amendment of the Labour Code

    Directory of Open Access Journals (Sweden)

    Jana Mervartová

    2012-01-01

    Full Text Available The amendment of the Labour Code, No. 365/2011 Coll., effective as from 1st January 2012, brings some of fundamental changes in labour law. The amendment regulates relation between the Labour Code and the Civil Code; and is also formulates principles of labour law relations newly. The basic period by fixed-term contract of employment is extended and also frequency its conclusion is limited. The length of trial period and the amount of redundancy payment are graduated. An earlier legislative regulation which an employee is temporarily assign to work for different employer has been returned. The number of hours by agreement to perform work is increased. The monetary compensation by competitive clause is reduced. The other changes are realised in part of collective labour law. The authoress of article notifies of the most important changes. She compares new changes of the Labour Code and former legal system and she also evaluates their advantages and disadvantages. The main objective of changes ensures labour law relations to be more flexible. And it should motivate creation of new jobs opening by employers. Amended provisions are aimed to reduction expenses of employers under the reform of the public finances. Also changes are expected in the Labour Code in connection with the further new Civil Code.