Sample records for on-line pwr rhr

  1. On-line PWR RHR pump performance testing following motor and impeller replacement

    DiMarzo, J.T.


    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  2. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail:, E-mail:, E-mail: [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear


    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  3. On-line measurement of gaseous iodine species during a PWR severe accident

    Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)


    A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

  4. Hygrometric measurement for on-line monitoring of PWR vessel head penetrations; Detection de fuites de traversees de couvercles de cuve par surveillance hygrometrique

    Germain, J.L.; Loisy, F.; Apolzan, S.


    In September 1991, a small leak was found on one of the reactor`s upper vessel head penetrations. After inspection, other non-throughwall cracks were localized in the lower part of the vessel head adapter in questions. The same type of crack was later found inside some adapters on other French PWR units. After repairs, the safety authorities granted approval to continue unit operation, with the specific provision that a system for ongoing monitoring of the penetrations be set up. Two types of system were selected to detect leaks through any potential cracks: the first is based on nitrogen-13 detection and the second on steam detection. Both systems call for sampling the air in a confined space above the vessel head. The number and distribution of sampling taps in the circuit, and the balancing of their respective flow rates, are factors in proper monitoring of all vessel head penetrations. Gas-injection holes are also installed in the confined space. These holes are used during the sampling system qualification tests to simulate leaks in various positions and calculate the effective performance of the sampling system. Leaks are simulated using a helium-base gas tracer and measuring tracer concentrations in the sampling system. The system for measuring steam levels in air samples uses chilled-mirror hygrometers. A microcomputer takes regular readings, drives the various automatic functions of the measurement system and automatically analyses the readings so as to monitor operations and trigger an alarm at the first sign of a leak. This system has now been installed for a year and a half on three French PWR units and is functioning satisfactorily. (authors). 5 figs.

  5. Molecular cloning, sequencing, and distribution of feline GnRH receptor (GnRHR) and resequencing of canine GnRHR.

    Samoylov, Alexandre M; Napier, India D; Morrison, Nancy E; Martin, Douglas R; Cox, Nancy R; Samoylova, Tatiana I


    GnRH receptors play vital roles in mammalian reproduction via regulation of gonadotropin secretion, which is essential for gametogenesis and production of gonadal steroids. GnRH receptors for more than 20 mammalian species have been sequenced, including human, mouse, and dog. This study reports the molecular cloning and sequencing of GnRH receptor (GnRHR) cDNA from the pituitary gland of the domestic cat, an important species in biomedical research. Feline GnRHR cDNA is composed of 981 nucleotides and encodes a 327 amino acid protein. Unlike the majority of mammalian species sequenced so far, but similar to canine GnRHR, feline GnRHR protein lacks asparagine in position three of the extracellular domain of the protein. At the amino acid level, feline GnRHR exhibits 95.1% identity with canine, 93.8% with human, and 88.9% with mouse GnRHR. Comparative sequence analysis of GnRHRs for multiple mammalian species led to resequencing of canine GnRHR, which differed from that previously published by a single base change that translates to a different amino acid in position 193. This single base change was confirmed in dogs of multiple breeds. Reverse transcriptase PCR analysis of GnRHR messenger RNA in different tissues from four normal cats indicated the presence of amplicons of varying lengths, including full-length as well as shortened GnRHR amplicons, pointing to the existence of truncated GnRHR transcripts in the domestic cat. This study is the first insight into molecular composition and expression of feline GnRHR and promotes better understanding of receptor organization, and distribution in various tissues of this species.

  6. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    HUO Xiao-Dong; XIE Zhong-Sheng


    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  7. Standard PWR for Italy

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))


    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  8. PWR decontamination feasibility study

    Silliman, P.L.


    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  9. Distribution Patterns and Developmental Changes of GnRH and GnRHR-Immunopositive Cells in the Pituitary of Ji Ning Gray Goats

    Liu Xiao, Wang Shu-Ying, Huang Li-Bo, Hou Yan-Meng and Shi Yun-Zhi


    Full Text Available Immunohistochemical Strept Avidin Biotin-Peroxidase Complex (SABC three-step method was used to investigate the distribution patterns and developmental changes of GnRH and GnRHR immunopositive (GnRH-ip and GnRHR-ip cells in the pituitary of Ji Ning Gray goats during 0-180 days of age. The results showed that GnRH-ip and GnRHR-ip cells were detected only in the pars distalis of adenohypophysis. There were no positive cells in the pars intermedia and neurohypophysis. GnRH-ip and GnRHR-ip cells were not observed in the anterior pituitary at birth day and 30 days of age. At 60 days, a number of GnRH-ip and GnRHR-ip cells were found in the anterior pituitary. GnRH-ip cells were pale brown which scattered or clustered around the sinusoid capillary; GnRHR-ip cells were brown and the cytomembrane was darker. The size and percentage of GnRH-ip and GnRHR-ip cells increased with the age growth. The numbers of GnRH-ip and GnRHR-ip cells after sexual maturity were significantly bigger than that before sexual maturity. The results above suggested that GnRH and GnRHR in the pituitary of Ji Ning Gray goats play a pivotal role in the sexual development and sexual maturity.

  10. Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

    Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)


    When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

  11. Physics of hydride fueled PWR

    Ganda, Francesco

    The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

  12. Study for identification of control rod drops in PWR reactors at any burnup step

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail:, E-mail:, E-mail:, E-mail: [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)


    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  13. Restoration of testis function in hypogonadotropic hypogonadal mice harboring a misfolded GnRHR mutant by pharmacoperone drug therapy.

    Janovick, Jo Ann; Stewart, M David; Jacob, Darla; Martin, L D; Deng, Jian Min; Stewart, C Allison; Wang, Ying; Cornea, Anda; Chavali, Lakshmi; Lopez, Suhujey; Mitalipov, Shoukhrat; Kang, Eunju; Lee, Hyo-Sang; Manna, Pulak R; Stocco, Douglas M; Behringer, Richard R; Conn, P Michael


    Mutations in receptors, ion channels, and enzymes are frequently recognized by the cellular quality control system as misfolded and retained in the endoplasmic reticulum (ER) or otherwise misrouted. Retention results in loss of function at the normal site of biological activity and disease. Pharmacoperones are target-specific small molecules that diffuse into cells and serve as folding templates that enable mutant proteins to pass the criteria of the quality control system and route to their physiologic site of action. Pharmacoperones of the gonadotropin releasing hormone receptor (GnRHR) have efficacy in cell culture systems, and their cellular and biochemical mechanisms of action are known. Here, we show the efficacy of a pharmacoperone drug in a small animal model, a knock-in mouse, expressing a mutant GnRHR. This recessive mutation (GnRHR E(90)K) causes hypogonadotropic hypogonadism (failed puberty associated with low or apulsatile luteinizing hormone) in both humans and in the mouse model described. We find that pulsatile pharmacoperone therapy restores E(90)K from ER retention to the plasma membrane, concurrently with responsiveness to the endogenous natural ligand, gonadotropin releasing hormone, and an agonist that is specific for the mutant. Spermatogenesis, proteins associated with steroid transport and steroidogenesis, and androgen levels were restored in mutant male mice following pharmacoperone therapy. These results show the efficacy of pharmacoperone therapy in vivo by using physiological, molecular, genetic, endocrine and biochemical markers and optimization of pulsatile administration. We expect that this newly appreciated approach of protein rescue will benefit other disorders sharing pathologies based on misrouting of misfolded protein mutants.

  14. Clinical assessment and molecular analysis of GnRHR and KAL1 genes in males with idiopathic hypogonadotrophic hypogonadism.

    Versiani, Beatriz R; Trarbach, Ericka; Koenigkam-Santos, Marcel; Dos Santos, Antonio Carlos; Elias, Lucila L K; Moreira, Ayrton C; Latronico, Ana Claudia; de Castro, Margaret


    The pathogenesis of idiopathic hypogonadotrophic hypogonadism (IHH) is mostly unclear. We characterized the clinical findings and molecular analysis of GnRHR and KAL1 genes in 26 Brazilian males with IHH with and without hyposmia/anosmia. Design Clinical assessment was performed for endocrine status, olfactory structure and function, renal lesion, and mirror movement. The diagnosis of Kallmann syndrome (KS) included HH and the clinical complaint of hyposmia/anosmia or decreased olfactory acuity obtained by the Smell Identification Test (SIT). We analysed GnRHR and KAL1 genes using the polymerase chain reaction (PCR) direct sequencing method. A variable degree of HH was observed, including various clinical abnormalities, such as cryptorchidism, hearing loss, strabismus, cleft lip/palate, high-arched palate, dental agenesis, psychiatric disorders, learning dysfunction, and bimanual synkinesia. Twenty-two out of 26 patients with IHH (85%) were classified as KS. Abnormalities of olfactory bulbs/sulci were observed in 79% of KS patients. One-third of KS patients had renal defects and 45.5% had a positive family history. GnRHR gene sequence analysis showed no mutations. KAL1 sequence analysis identified two novel missense mutations: c.1061A to G in exon 7 (N304S) and c.1583C to A in exon 10 (S478X). We also observed a 14-bp deletion within exon 11 that caused a premature termination. According to the National Center for Biotechnology Information (NCBI)-Single Nucleotide Polymorphism (SNP) database, two previously described polymorphisms (rs808119 and rs809446) were detected. KAL1 mutations accounted for 12% of KS patients. This low prevalence of KAL1 mutations indicates that other genes, such as the fibroblast growth factor receptor 1 (FGFR1) gene or other as yet undiscovered genes, epigenetic events and/or environmental factors might be involved in the aetiology and phenotypic variability of KS.

  15. Effect of short-term and prolonged stress on the biosynthesis of gonadotropin-releasing hormone (GnRH) and GnRH receptor (GnRHR) in the hypothalamus and GnRHR in the pituitary of ewes during various physiological states.

    Ciechanowska, M; Łapot, M; Antkowiak, B; Mateusiak, K; Paruszewska, E; Malewski, T; Paluch, M; Przekop, F


    Using an ELISA assay, the levels of GnRH and GnRHR were analysed in the preoptic area (POA), anterior (AH) and ventromedial hypothalamus (VM), stalk/median eminence (SME); and GnRHR in the anterior pituitary gland (AP) of non-breeding and breeding sheep subjected to short-term or prolonged stress. The ELISA study was supplemented with an analysis of plasma LH concentration. Short-term footshock stimulation significantly increased GnRH levels in hypothalamus in both seasons. Prolonged stress elevated or decreased GnRH concentrations in the POA and the VM, respectively during anoestrus, and lowered GnRH amount in the POA-hypothalamus of follicular-phase sheep. An up-regulation of GnRHR levels was noted in both, anoestrous and follicular-phase animals. In the non-breeding period, a prolonged stress procedure increased GnRHR biosynthesis in the VM and decreased it in the SME and AP, while in the breeding time the quantities of GnRHR were significantly lower in the whole hypothalamus. In follicular-phase ewes the fluctuations of GnRH and GnRHR levels under short-term and prolonged stress were reflected in the changes of LH secretion, suggesting the existence of a direct relationship between GnRH and GnRH-R biosynthesis and GnRH/LH release in this period. The study showed that stress was capable of modulating the biosynthesis of GnRH and GnRHR; the pattern of changes was dependent upon the animal's physiological state and on the time course of stressor application. The obtained results indicate that the disturbances of gonadotropin secretion under stress conditions in sheep may be due to a dysfunction of GnRH and GnRHR biosynthetic pathways. Copyright © 2016 Elsevier B.V. All rights reserved.

  16. Diagnosis and processing of detector failure in PWR on-line power distribution monitoring%压水堆堆芯功率分布在线监测计算中的探测器失效诊断与处理方法

    李茁; 吴宏春; 曹良志; 李云召; 刘宙宇


    Measurements of in-core detectors,as one of the most important inputs for a reactor core on-line power distribu-tion monitoring system,seriously affect the on-line monitoring results.Therefore,the diagnosis and processing of detector failure are necessary.Harmonics expansion method is employed to the on-line monitoring calculation of reactor core power distribution. To diagnose failure of detectors,three methods— direct method,comparison of measurements method and comparison of recon-structed responses method—are used in combination.Based on these methods,diagnosis and processing of detector failure have been added into NECP-ONION,an in-house online monitoring code developed by NECP Lab.Benchmark for Evaluation and Vali-dation Reactor Simulations(BEAVRS)has been used for code validation,especially for the diagnosis and processing of detector failure.Numerical results show that,the combination of these three methods is not only effective to the complete failed condition, but also useful for measurements deviation of normal value.In addition,NECP-ONION is functional to distinguish detector fail-ure or local power oscillation.This could reduce the misdiagnosis for detector failure.For the processing of detector failure,it has been observed that single detector failure or a small number of detectors failed has limit effect on the monitoring system NECP-ONION.%作为堆芯功率分布在线监测计算的重要输入参数,堆内中子探测器的测量数据对堆芯功率分布在线监测具有重要影响,因此在线监测计算中探测器失效的诊断和处理十分必要。采用谐波展开法进行堆芯功率分布的在线监测计算,采用直接观察法、探测器测量值比较法以及探测器重构值比较法分三个阶段对探测器失效进行诊断。基于以上理论,在在线监测系统 NECP-ONION中加入探测器失效诊断功能,利用BEAVRS基准题对失效诊断和处理进行验证。数值结果表明,将三阶段的诊断

  17. Neutron noise measurements on Bugey 2 PWR

    Marini, J.; Romy, D.; Spadi, J.C.; Assedo, R.; Castello, G.


    Following Bugey 2 PWR hot functional tests, dimension measurements of internals hold down spring led to suspect that vibration levels could change with time. Neutron noise measurements runs during the first cycle enabled describing vibration behaviour of internals. Comparisons with previous analytical and experimental results gained on the Safran model as well as on similar reactors were also made.

  18. Effects of MboII and BspMI polymorphisms in the gonadotropin releasing hormone receptor (GnRHR) gene on sperm quality in Holstein bulls.

    Yang, Wu-Cai; Tang, Ke-Qiong; Yu, Jun-Na; Zhang, Chun-Yan; Zhang, Xiao-Xia; Yang, Li-Guo


    The hypothalamic gonadotropin-releasing hormone receptor (GnRHR) plays an essential physiological role in reproductive function, which triggers the synthesis and release of luteinizing hormone and follicle stimulating hormone in the pituitary. The objective of this study was to investigate the effects of polymorphisms of GnRHR gene on the quality of fresh and frozen semen in Holstein bulls. The PCR-RFLP method was applied to detect G286A and T340C transitions determining MboII and BspMI polymorphisms, respectively, in the exon I of bovine GnRHR gene and evaluated its associations with sperm quality traits in 131 Holstein bulls. In polymorphic locus 286, bulls with the GA genotype had significantly higher sperm motility in frozen semen (FMOT) than GG genotype (P bulls with heterozygote CT genotype had significantly higher sperm motility (MOT), semen volume per ejaculate (VOL), and lower abnormal spermatozoa rate (ASR) than homozygote TT genotype (P Bulls contained one A allele or C allele had a favorable, positive effect on sperm quality traits. These results indicate that GnRHR gene can be a potential marker for improving sperm quality traits, and imply that bulls with GA or CT genotype should be selected in breeding program.

  19. Associations of Gonadotropin-Releasing Hormone Receptor (GnRHR) and Neuropeptide Y (NPY) Genes' Polymorphisms with Egg-Laying Traits in Wenchang Chicken

    WU Xu; ZHU Wen-qi; LI Hui-fang; YAN Mei-jiao; TANG Qing-ping; CHEN Kuan-wei; WANG Jin-yu; GAO Yu-shi; TU Yun-jie; YU Ya-bo


    Single nucleotide polymorphisms (SNP) of chicken gonadotropin-releasing hormone receptor (GnRHR) and neuropeptide Y (NPY) were selected to identify the genotypes of Wenchang (Chinese indigenous breed) chicken with restricton fragment length polymorphisms. The associations of the SNPs with the total egg production (NE), average days of continual laying (ADCL), and number of double-yolked eggs (DYE) traits were analyzed. The frequency of restriction enzyme A/a alleles in the population was for GnRHR 0.69 (Bpu1102 Ⅰ A) and 0.31 (Bpu1102 Ⅰ a) and for NPY 0.46 (Dra Ⅰ B) and 0.54 (Dra Ⅰ b). Trait data from a total of 120 hens, which were purebred introduced from Hainan Province, China from one generation were recorded. Two significant effects of genes' marker were found: for GnRHR and number of eggs (dominant; t= 2.67, df= 116) and NPY and number of eggs (additive; t= 1.97, df= 116). The current research supports the effects of GnRHR and NPY genes on egg-laying traits of chickens.

  20. APT Blanket Safety Analysis: Preliminary Analyses of Downflow Through a Lateral Row 1 Blanket Model Under Near RHR Conditions

    Hamm, L.L.


    To address a concern about a potential maldistribution of coolant flow through an APT blanket module under low flow near RHR conditions, a scoping study of downflow mixed convection in parallel channels was conducted. Buoyancy will adversely effect the flow distribution in module bins with downflow and non-uniform power distributions. This study consists of two parts: a simple analytical model of flow in a two channel network, and a lumped eleven channel FLOWTRAN-TF model of a front lateral Row-1 blanket module bin. Results from both models indicate that the concern about coolant flow in a vertical model being diverted away from high power regions by buoyancy is warranted. The FLOWTRAN-TF model predicted upflow (i.e., a flow reversal) through several of the high power channels, under some low flow conditions. The transition from the regime with downflow in all channels to a regime with upflow in some channels was abrupt.

  1. Aberrant gonadotropin-releasing hormone receptor (GnRHR) expression and its regulation of CYP11B2 expression and aldosterone production in adrenal aldosterone-producing adenoma (APA).

    Nakamura, Yasuhiro; Hattangady, Namita G; Ye, Ping; Satoh, Fumitoshi; Morimoto, Ryo; Ito-Saito, Takako; Sugawara, Akira; Ohba, Koji; Takahashi, Kazuhiro; Rainey, William E; Sasano, Hironobu


    Aberrant expression of gonadotropin-releasing hormone receptor (GnRHR) has been reported in human adrenal tissues including aldosterone-producing adenoma (APA). However, the details of its expression and functional role in adrenals are still not clear. In this study, quantitative RT-PCR analysis revealed the mean level of GnRHR mRNA was significantly higher in APAs than in human normal adrenal (NA) (P=0.004). GnRHR protein expression was detected in human NA and neoplastic adrenal tissues. In H295R cells transfected with GnRHR, treatment with GnRH resulted in a concentration-dependent increase in CYP11B2 reporter activity. Chronic activation of GnRHR with GnRH (100nM), in a cell line with doxycycline-inducible GnRHR (H295R-TR/GnRHR), increased CYP11B2 expression and aldosterone production. These agonistic effects were inhibited by blockers for the calcium signaling pathway, KN93 and calmidazolium. These results suggest GnRH, through heterotopic expression of its receptor, may be a potential regulator of CYP11B2 expression levels in some cases of APA. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  2. Shielding design for PWR in France

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.


    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983.

  3. The integrated PWR; Les REP integres

    Gautier, G.M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Reacteurs


    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  4. On-line measurements of RuO{sub 4} during a PWR severe accident

    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)


    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  5. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Tanner, V.; Rosenberg, R. [VTT Chemical Technology, Otaniemi (Finland)


    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  6. Alterations to dendritic spine morphology, but not dendrite patterning, of cortical projection neurons in Tc1 and Ts1Rhr mouse models of Down syndrome.

    Matilda A Haas

    Full Text Available Down Syndrome (DS is a highly prevalent developmental disorder, affecting 1/700 births. Intellectual disability, which affects learning and memory, is present in all cases and is reflected by below average IQ. We sought to determine whether defective morphology and connectivity in neurons of the cerebral cortex may underlie the cognitive deficits that have been described in two mouse models of DS, the Tc1 and Ts1Rhr mouse lines. We utilised in utero electroporation to label a cohort of future upper layer projection neurons in the cerebral cortex of developing mouse embryos with GFP, and then examined neuronal positioning and morphology in early adulthood, which revealed no alterations in cortical layer position or morphology in either Tc1 or Ts1Rhr mouse cortex. The number of dendrites, as well as dendrite length and branching was normal in both DS models, compared with wildtype controls. The sites of projection neuron synaptic inputs, dendritic spines, were analysed in Tc1 and Ts1Rhr cortex at three weeks and three months after birth, and significant changes in spine morphology were observed in both mouse lines. Ts1Rhr mice had significantly fewer thin spines at three weeks of age. At three months of age Tc1 mice had significantly fewer mushroom spines--the morphology associated with established synaptic inputs and learning and memory. The decrease in mushroom spines was accompanied by a significant increase in the number of stubby spines. This data suggests that dendritic spine abnormalities may be a more important contributor to cognitive deficits in DS models, rather than overall neuronal architecture defects.

  7. Horizontal Drop of 21- PWR Waste Package

    A.K. Scheider


    The objective of this calculation is to determine the structural response of the waste package (WP) dropped horizontally from a specified height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 16). AP-3.12Q, ''Calculations'' (Ref. 11) is used to perform the calculation and develop the document. The sketches attached to this calculation provide the potential dimensions and materials for the 21-PWR WP design.

  8. Study of safety relief valve operation under ATWS conditions. [PWR

    Hutmacher, E.S.; Nesmith, B.J.; Brukiewa, J.B.


    A literature survey and analysis project has been performed to determine if recent (since mid-1975) data has been reported which could influence the current approach to predicting PWR relief valve capacity under ATWS conditions. This study was conducted by the Energy Technology Engineering Center for NRC. Results indicate that the current relief valve capacity model tends to predict less capacity than actually obtains; however, no experimental verification at PWR ATWS conditions was found. Other project objectives were to establish the availability of methods for evaluating reaction forces and back pressure effects on relief valve capacity, and to determine if facilities exist which are capable of testing PWR relief valves at ATWS conditions.

  9. Characterization of Factors affecting IASCC of PWR Core Internals

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  10. The PWR cores management; La gestion des coeurs REP

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others


    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  11. Zebra: An advanced PWR lattice code

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)


    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  12. Degraded core analysis for the PWR

    Gittus, J.H.


    The paper presents an analysis of the probability and consequences of degraded core accidents for the PWR. The article is based on a paper which was presented by the author to the Sizewell-B public inquiry. Degraded core accidents are examined with respect to:- the initiating events, safety plant failure, and processes with a bearing on containment failure. Accident types and frequencies are discussed, as well as the dispersion of radionuclides. Accident risks, i.e. individual and societal risks in degraded core accidents are assessed from:- the amount of radionuclides released, the weather, the population distribution, and the accident frequencies. Uncertainties in the assessment of degraded core accidents are also summarized. (U.K.).

  13. A pressure drop model for PWR grids

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  14. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Yuann, Yng-Ruey, E-mail:


    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  15. Formación on line On line learning

    O. Grau-Perejoan


    Full Text Available La formación on line es una modalidad de enseñanza a distancia basada en las nuevas tecnologías. En este artículo se pretende hacer una introducción a base de describir a grandes rasgos sus características principales: asincronía, no presencialidad, comunicación escrita, función del profesor on line, así como los retos, los riesgos, las ventajas y los inconvenientes que plantea. Se exponen las diferencias entre la formación on line y la formación presencial, de manera que los docentes puedan adaptar de la mejor manera posible sus propuestas formativas a la modalidad on line. Se introduce el importantísimo papel de la planificación y de la fase de diseño y, finalmente, se repasan conceptos útiles para comprender mejor el mundo de la formación on line como son los conceptos entorno virtual de aprendizaje (EVA o Blended Learning (B-Learning.On line learning is a type of distance education based on new technologies. This article's aim is to introduce its main characteristics -asynchrony, non-presentiality, written communication, e-teacher role- as well as its challenges, risks, advantages and limitations. Differences between on line learning and face-to-face learning are presented in order to enable educational professionals to adapt their courses to the on line methodology. Planning and designing are introduced as key phases and, finally, useful concepts such as Virtual Learning Environment (VLE or Blended Learning (B-Learning are reviewed in order to achieve a better understanding of the on line learning field.

  16. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)


    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  17. Conceptual study on advanced PWR system

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.


    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  18. A PWR Thorium Pin Cell Burnup Benchmark

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.


    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  19. Conceptual study of advanced PWR core design

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong


    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  20. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Kim, Kyu-Tae, E-mail:


    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  1. Seismic qualification of PWR plant auxiliary feedwater systems

    Lu, S.C.; Tsai, N.C.


    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14.

  2. The advanced main control console for next japanese PWR plants

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)


    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  3. Evaluation of PWR and BWR pin cell benchmark results

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Unit Nuclear Energy, Netherlands Energy Research Foundation ECN, Petten (Netherlands)); Hoogenboorm, J.E.; De Leege, P.F.A. (International Reactor Institute IRI, University of Leiden, Leiden (Netherlands)); Van de Voet, J.; Verhagen, F.C.M. (KEMA NV, Arnhem (Netherlands))


    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary.

  4. Monte Carlo based radial shield design of typical PWR reactor

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.


    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  5. Leak before break application in French PWR plants under operation

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)


    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  6. Advanced ion exchange resins for PWR condensate polishing

    Hoffman, B. [Rohm and Haas Co. (United States); Tsuzuki, S. [Rohm and Haas Co. (Japan)


    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  7. Evaluation of PWR and BWR pin cell benchmark results

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))


    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  8. Evaluation of alternative descriptions of PWR cladding corrosion behavior

    Quecedo, M.; Serna, J. J.; Weiner, R. A.; Kersting, P. J.


    A statistical procedure has been used to evaluate several alternative descriptions of pressurized water reactor (PWR) cladding corrosion behavior, using an extensive database of Improved (low tin) Zr-4 cladding corrosion measurements from fuel irradiated in commercial PWRs. The in-reactor corrosion enhancement factors considered in the model development are based on a comprehensive review of the current literature for PWR cladding corrosion phenomenology and models. In addition, because prediction of PWR cladding corrosion behavior is very sensitive to the values used for the oxide surface temperatures, several models for the forced convection and sub-cooled nucleate boiling (SNB) coolant heat transfer under PWR conditions have also been evaluated. This evaluation determined that the choice of the forced convection heat transfer has the greatest impact on the ability to fit the data. In addition, the SNB heat transfer model used must account for a continuous transition from forced convection conditions to fully developed SNB conditions. With these choices for the heat transfer models, the evaluation determined that the significant in-reactor corrosion enhancement factors are related to the formation of a hydride rim at the cladding outer diameter, the coolant lithium concentration, and the fast neutron fluence (author) (ml)

  9. Studies of a small PWR for onsite industrial power

    Klepper, O.H.; Smith, W.R.


    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application.

  10. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Virpi Kouhia


    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  11. PWR fuel in Japan; The changes and trend for hereafter

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki


    As for the PWR fuel in Japan, much efforts have been exerted aiming at the high reliability since the start of operation of Mihama No. 1 plant of Kansai Electric Power Co., Inc. At the beginning of 1970s, the fuel made by Westinghouse in USA was imported, and since then, the pursuit of the causes of troubles and the countermeasures and the domestic production of fuel have been carried out, and the improvement of design and the strengthening of quality control have been advanced. As the results, the occurrence of troubles decreased rapidly. As the fuel improvement for hereafter, the economical improvement by higher burnup, the saving and effective use of uranium resources as well as the increase of reliability are emphasized. The changes in the PWR fuel by Westinghouse, the course of improvement in the PWR fuel in Japan, the improvement against the troubles of the fuel, the improved design, the verification of the performance of the PWR fuel, the trend of development of the fuel such as the heightening of burnup, the saving and effective use of uranium resources, and the improved type pressurized water reactors are reported. (K.I.).

  12. A neutronic study of the cycle PWR-CANDU

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail:;;;;


    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  13. CERN Video News on line


    The latest CERN video news is on line. In this issue : an interview with the Director General and reports on the new home for the DELPHI barrel and the CERN firemen's spectacular training programme. There's also a vintage video news clip from 1954. See: or Bulletin web page

  14. Methodology for the LABIHS PWR simulator modernization

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail:, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)


    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  15. Clean Air OnLine

    Finney, D. [Environment Canada, Gatineau, PQ (Canada). Air Pollution Prevention Directorate


    This presentation describes Clean Air OnLine, a multi-tiered website dedicated to providing Canadians with information on air quality. The website is under development to support action to reduce air emissions, demonstrate the links between air emissions and environmental impacts, and enhance the understanding of sustainable community development issues such as health, energy, and urban sprawl. Partners in the Clean Air OnLine (CAOL) initiative include Environment Canada and the Clean Air Partnership which includes the Greater Toronto Area pilot project. The audience for CAOL includes municipal decision makers, local decision makers, community leaders, and the general public. The project provides Canadians with air pollution contextual information on pollution sources, pollutants, and related issues. It also provides information on health, environmental and economic impacts and the interrelationships with climate change issues and energy use. tabs., figs.

  16. On-line moisture analysis

    Cutmore, N G


    Measurement of the moisture content of iron ore has become a key issue for controlling moisture additions for dust suppression. In most cases moisture content is still determined by manual or automatic sampling of the ore stream, followed by conventional laboratory analysis by oven drying. Although this procedure enables the moisture content to be routinely monitored, it is too slow for control purposes. This has generated renewed interest in on-line techniques for the accurate and rapid measurement of moisture in iron ore on conveyors. Microwave transmission techniques have emerged over the past 40 years as the dominant technology for on-line measurement of moisture in bulk materials, including iron ores. Alternative technologies have their limitations. Infra-red analysers are used in a variety of process industries, but rely on the measurement of absorption by moisture in a very thin surface layer. Consequently such probes may be compromised by particle size effects and biased presentation of the bulk mater...

  17. Assessment of PWR plutonium burners for nuclear energy centers

    Frankel, A J; Shapiro, N L


    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  18. PWR fuel in Japan; Progress and future trends

    Yokote, Mitsuhiro (Kansai Electric Power Co., Inc., Osaka (Japan)); Kondo, Yoshiaki; Abeta, Sadaaki (Mitsubishi Heavy Industries Ltd., Tokyo (Japan))


    Twenty years ago, in the early years of the Japanese civil nuclear power programme, the fuel used was imported from Westinghouse in the USA. However, it was always intended that there would be a move towards fuel fabrication in Japan and by the end of 1993 around 10,000 Mitsubishi PWR fuel assemblies had been supplied to 21 PWRs in Japan. The highest burnup achieved so far is 46 GWd/t. Design changes to reduce abnormalities have been made, reliability is improving all the time and further improvements in burnup are being developed. This progress in PWR cores and fuel including MOX fuel in Japan is charted and future research and development is outlined. (UK).

  19. A concept of PWR using plate and shell heat exchangers

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail:, E-mail: [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)


    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  20. Control of corrosion product transport in PWR secondary cycles

    Sawochka, S.G.; Pearl, W.L. [NWT Corp., San Josa, CA (United States); Passell, T.O.; Welty, C.S. [Electric Power Research Institute, Palo Alto, CA (United States)


    Transport of corrosion products to PWR steam generators by the feedwater leads to sludge buildup on the tubesheets and fouling of tube-to-tube support crevices. In these regions, chemical impurities concentrate and accelerate tubing corrosion. Deposit buildup on the tubes also can lead to power generation limitations and necessitate chemical cleaning. Extensive corrosion product transport data for PWR secondary cycles has been developed employing integrating sampling techniques which facilitate identification of major corrosion product sources and assessments of the effectiveness of various control options. Plant data currently are available for assessing the impact of factors such as pH, pH control additive, materials of construction, blowdown, condensate treatment, and high temperature drains and feedwater filtration.

  1. Evaluation of PWR and BWR pin cell benchmark results

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))


    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  2. Study on thermal-hydraulics during a PWR reflood phase

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  3. PHENIX on-line systems

    Adler, S.S.; Allen, M.; Alley, G.; Amirikas, R.; Arai, Y.; Awes, T.C.; Barish, K.N.; Barta, F.; Batsouli, S.; Belikov, S.; Bennett, M.J.; Bobrek, M.; Boissevain, J.G.; Boose, S.; Britton, C.; Britton, L.; Bryan, W.L.; Cafferty, M.M.; Carey, T.A.; Chang, W.C.; Chi, C.Y.; Chiu, M.; Cianciolo, V.; Cole, B.A.; Constantin, P.; Cook, K.C.; Cunitz, H.; Desmond, E.J.; Ebisu, K.; Efremenko, Y.V.; El Chenawi, K.; Emery, M.S.; Engo, D.; Ericson, N.; Fields, D.E.; Frank, S.; Frantz, J.E.; Franz, A.; Frawley, A.D.; Fried, J.; Gannon, J.; Gee, T.F.; Gentry, R.; Giannotti, P.; Gustafsson, H.-A.; Haggerty, J.S.; Hahn, S.; Halliwell, J.; Hamagaki, H.; Hansen, A.G.; Hara, H.; Harder, J.; He, X.; Heistermann, F.; Hemmick, T.K.; Hibino, M.; Hill, J.C. E-mail:; Homma, K.; Jacak, B.V.; Jagadish, U.; Jia, J.; Kajihara, F.; Kametani, S.; Kamyshkov, Y.; Kandasamy, A.; Kang, J.H.; Kapustinsky, J.; Katou, K.; Kelley, M.A.; Kelly, S.; Kikuchi, J.; Kim, S.Y.; Kim, Y.G.; Kistenev, E.; Kotchetkov, D.; Kurita, K.; Lajoie, J.G.; Lenz, M.; Lenz, W.; Li, X.H.; Lin, S.; Liu, M.X.; Markacs, S.; Matathias, F.; Matsumoto, T.; Mead, J.; Mischke, R.E.; Mishra, G.C.; Moore, A.; Muniruzzamann, M.; Musrock, M.; Nagle, J.L.; Nandi, B.K.; Newby, J.; Nystrand, J.; O' Brien, E.; O' Connor, P.; Ohnishi, H.; Oskarsson, A.; Osterman, L.; Oyama, K.; Paffrath, L.; Pancake, C.E.; Pantuev, V.S.; Petridis, A.N.; Pisani, R.P.; Plagge, T.; Plasil, F.; Purschke, M.L.; Rankowitz, S.; Rao, R.; Rau, M.; Read, K.F.; Ryu, S.S.; Sakaguchi, T.; Sato, H.D.; Seto, R.; Shiina, T.; Silvermyr, D.; Simon-Gillo, J.; Simpson, M.; Sippach, W.; Skank, H.D.; Skutnik, S.; Sleege, G.A.; Smith, G.D.; Smith, M.; Stankus, P.W.; Steinberg, P.; Sugitate, T.; Sullivan, J.P.; Taketani, A.; Tamai, M.; Tanaka, Y.; Thomas, W.D.; Todd, R.; Toldo, F.; Turner, G.; Ushiroda, T.; Velkovska, J.; Hecke, H.W. van; Lith, M. van; Villatte, L.; Achen, W. von; Walker, J.W.; Wang, H.Q.; White, S.N.; Wintenberg, A.L.; Witzig, C.; Wood, L.[and others


    The PHENIX On-Line system takes signals from the Front End Modules (FEM) on each detector subsystem for the purpose of generating events for physics analysis. Processing of event data begins when the Data Collection Modules (DCM) receive data via fiber-optic links from the FEMs. The DCMs format and zero suppress the data and generate data packets. These packets go to the Event Builders (EvB) that assemble the events in final form. The Level-1 trigger (LVL1) generates a decision for each beam crossing and eliminates uninteresting events. The FEMs carry out all detector processing of the data so that it is delivered to the DCMs using a standard format. The FEMs also provide buffering for LVL1 trigger processing and DCM data collection. This is carried out using an architecture that is pipelined and deadtimeless. All of this is controlled by the Master Timing System (MTS) that distributes the RHIC clocks. A Level-2 trigger (LVL2) gives additional discrimination. A description of the components and operation of the PHENIX On-Line system is given and the solution to a number of electronic infrastructure problems are discussed.

  4. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Thiollay Nicolas


    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  5. PWR Cross Section Libraries for ORIGEN-ARP

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL


    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

  6. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad


    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  7. Validation of gadolinium burnout using PWR benchmark specification

    Oettingen, Mikołaj, E-mail:; Cetnar, Jerzy, E-mail:


    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  8. PWR core stablity aganst xenon-induced spatial power oscillation

    Moon, H.J.; Han, K.I. (Korea Advanced Energy Research Inst., Seoul (Republic of Korea))


    Stability of a PWR core against xenon-induced axial power oscillation is studied using one-dimensional xenon transient analysis code, DD1D, that has been developed and verified at KAERI. Analyzed by DD1D utilizing the Kori Unit 1 design and operating data is the sensitivity of axial stability in a PWR core to the changes in core physical parameters including core power level, moderator temperature coefficient, core inlet temperature, doppler power coefficient and core average burnup. Through the sensitivity study the Kori Unit 1 core is found to be stable against axial xenon oscillation at the beginning of cycle 1. But, it becomes less stable as burnup progresses, and unstable at the end of cycle. Such a decrease in stability is mainly due to combined effect of changes in axial power distribution, moderator temperature coefficient and doppler power coefficient as core burnup progresses. It is concluded from the stability analysis of the Kori Unit 1 core that design of a large PWR with high power density and increased dimension can not avoid xenon-induced axial power instabilites to some extents, especially at the end of cycle.

  9. Actinides transmutation - a comparison of results for PWR benchmark

    Claro, Luiz H. [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)], e-mail:


    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO{sub 2} used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k{infinity} and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  10. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Gomez Briceno, D.; Serrano, M.


    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)


    Pande Made Udiyani


    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  12. EPRI PWR Safety and Relief Valve Test Program: test condition justification report

    Hosler, J.


    In response to NUREG 0737, Item II.D.1.A requirements, several safety and relief valve designs were tested by EPRI under PWR utility sponsorship. Justification that the inlet fluid conditions under which these valve designs were tested are representative of those expected in participating domestic PWR units during FSAR, Extended High Pressure Injection, and Cold Overpressurization events is presented.

  13. PWR safety and relief valve test program. Valve selection/juftification report. Final report


    NUREG 0578 required that full-scale testing be performed on pressurizer safety valves and relief valves representative of those in use or planned for use in PWR plants. To obtain valve performance data for the entire population of PWR plant valves, nine safety valves and ten relief valves were selected as a fully representative set of test valves. Justification that the selected valves represent all PWR plant valves was provided by each safety and relief valve manufacturer. Both the valve selection and justification work was performed as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of the PWR utilities in response to the recommendations of NUREG 0578 and the requirements of the NRC. Results of the Safety and Relief Valve Selection and Justification effort is documented in this report.

  14. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Jimenez, G.; Matias, R.; Fernandez, K.; Justo, D.; Bocanegra, R.; Mena, L.; Queral, C.


    During a severe accident in a PWR, the hydrogen generated may be distributed in the containment atmosphere and reach the combustion conditions that can cause the containment failure. In this research project, a preliminary study has been done about the capacities of ANSYS Fluent 15.0 and GOTHIC 8.0 to tri dimensional distribution of the hydrogen in a PWR containment during a severe accident. (Author)

  15. A study on thimble plug removal for PWR plants

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)


    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  16. Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

    S. Mastilovic; A. Scheider; S.M. Bennett


    The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.

  17. Estimating probable flaw distributions in PWR steam generator tubes

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)


    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses.

  18. Integral Test Facility PKL: Experimental PWR Accident Investigation


    Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circul...

  19. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Kaercher, S.


    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  20. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Esteban, G.


    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  1. VERA Core Simulator Methodology for PWR Cycle Depletion

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL


    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  2. SCOR 1000: an economic and innovative conceptual design PWR

    Gautier, G.M.; Chenaud, M.S. [CEA Cadarache (DEN/DER/SESI), 13 - Saint Paul lez Durance (France). Dept. d' Etudes des Reacteurs; Tourniaire, B. [CEA Grenoble (DEN/DTN/SE2T/LPTM), 38 (France)


    Within the framework of innovative reactors studies, the Cea proposes the SCOR design (Simple COmpact Reactor) based on most of the advantages of innovative reactors. All main components are integrated in the vessel: the pressurizer, the canned pumps, the control rod mechanics of the driving system (CMD), and the dedicated heat exchangers of the passive heat removal system. The only steam generator is located above the vessel instead of the upper head. This design is featured by its compactness and by a large suppression or simplification of auxiliary systems. The first design with a 600 MWe shows its competitiveness with regard to the large loop-type PWR. To reduce the cost investment by the law sized effect, we examine the possibility of increasing the power of the reactor, while keeping the safety advantages of the medium sized SCOR. The electrical power of the new design is 1000 MWe. SCOR-1000 operates at much lower primary circuit pressure than standard PWRs (93 bars instead of the usual 155 bars), and the power density is lower (80 MW/m3 instead of 100 for the present PWRs). The reactivity is controlled by the CMD and by the burnable poison, without soluble boron. With the same safety advantages of the medium-sized SCOR, the cost reduction of the investment and of cost production could reach 18% with regard to the loop-type PWR. (authors)

  3. PWR fuel performance and burnup extension in Japan

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.


    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  4. Degradation of fastener in reactor internal of PWR

    Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H


    Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

  5. PWR reactor vessel in-service-inspection according to RSEM

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul [Intercontrole, 13, rue du Capricorne - SILIC 433, 94583 Rungis - Cedex (France)


    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high

  6. Are on-line currencies virtual banknotes?

    Stephen F. Quinn; William Roberds


    The history of money is marked by innovations that have expanded the role of "inside money"-money created by the private sector. For instance, the past few years have seen the development of several types of on-line payment arrangements, some of which have been dubbed "on-line currencies." ; This article examines the likely success or failure of on-line currencies by means of a historical analogy. The discussion compares the introduction of on-line currencies to the debut of the bearer bankno...

  7. On line routing per mobile phone

    Bieding, Thomas; Görtz, Simon; Klose, Andreas


    On-line routing is concerned with building vehicle routes in an ongoing fashion in such a way that customer requests arriving dynamically in time are efficiently and effectively served. An indispensable prerequisite for applying on-line routing methods is mobile communication technology. Addition...

  8. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Fang, Jun, E-mail: [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)


    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  9. Modeling local chemistry in PWR steam generator crevices

    Millett, P.J. [EPRI, Palo Alto, CA (United States)


    Over the past two decades steam generator corrosion damage has been a major cost impact to PWR owners. Crevices and occluded regions create thermal-hydraulic conditions where aggressive impurities can become highly concentrated, promoting localized corrosion of the tubing and support structure materials. The type of corrosion varies depending on the local conditions, with stress corrosion cracking being the phenomenon of most current concern. A major goal of the EPRI research in this area has been to develop models of the concentration process and resulting crevice chemistry conditions. These models may then be used to predict crevice chemistry based on knowledge of bulk chemistry, thereby allowing the operator to control corrosion damage. Rigorous deterministic models have not yet been developed; however, empirical approaches have shown promise and are reflected in current versions of the industry-developed secondary water chemistry guidelines.

  10. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)


    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  11. PWR steam generator chemical cleaning, Phase I. Final report

    Rothstein, S.


    United Nuclear Industries (UNI) entered into a subcontract with Consolidated Edison Company of New York (Con Ed) on August 8, 1977, for the purpose of developing methods to chemically clean the secondary side tube to tube support crevices of the steam generators of Indian Point Nos. 1 and 2 PWR plants. This document represents the first reporting on activities performed for Phase I of this effort. Specifically, this report contains the results of a literature search performed by UNI for the purpose of determining state-of-the-art chemical solvents and methods for decontaminating nuclear reactor steam generators. The results of the search sought to accomplish two objectives: (1) identify solvents beyond those proposed at present by UNI and Con Ed for the test program, and (2) confirm the appropriateness of solvents and methods of decontamination currently in use by UNI.

  12. PWR and BWR spent fuel assembly gamma spectra measurements

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.


    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  13. Identifying thermal cycling mechanisms in PWR branch line piping

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Keller, J.D.; Bilanin, A.J. [Continuum Dynamics, Inc., Ewing, NJ (United States)


    Predicting the onset and the characteristics of thermal cycling in pressurized water reactor (PWR) branch line piping systems is critical to formulation of thermal fatigue screening tools. The complex nature of the underlying thermal-hydraulic phenomena, however, significantly complicates prediction using analytical models or direct numerical simulations. Instead, it is necessary to perform scaled experiments to identify the physical mechanisms and to gather data for formulation of semi-empirical models for the thermal cycling phenomena. Through the EPRI Materials Reliability Program a test program is underway to identify and develop semi-empirical correlations for the physical thermalhydraulic mechanisms that cause thermal cycling in dead-ended PWR branch line piping systems. Three series of tests are being performed in this test program: configuration tests on a representative up-horizontal (UH) branch line piping geometry, configuration tests on a representative down-horizontal (DH) branch line piping geometry, and high Reynolds number tests to assess penetration of secondary flow structures into a dead-ended branch line. Results from UH and DH configuration tests indicate that random turbulence penetration is not sufficient for thermal cycling to occur. Rather a swirling flow structure, representative of a large, 'corkscrew' vortical structure, is required for thermal cycling. Scale tests on the UH configuration have simulated cycling phenomena observed in full-scale plant data and have been used to determine parametric sensitivities in formulating a predictive model for the thermal cycling. Data indicate that the mechanism for thermal cycling in UH configurations is stochastic but scales with the leak rate from the valve. The critical dependent variables are reduced to several non-dimensional scaling curves, resulting in a semiempirical predictive model. This paper discusses the test program and the results obtained to date. Application of these

  14. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    M.Krug, R.Shogan


    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  15. Identification and evaluation of PWR in-vessel severe accident management strategies

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)


    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  16. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    M.Krug, R.Shogan


    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR.

  17. On-line generalized Steiner problem

    Awerbuch, B.; Azar, Y.; Bartal, Y. [Tel Aviv Univ. (Israel)


    The Generalized Steiner Problem (GSP) is defined as follows. We are given a graph with non-negative weights and a set of pairs of vertices. The algorithm has to construct minimum weight subgraph such that the two nodes of each pair are connected by a path. We consider the on-line generalized Steiner problem, in which pairs of vertices arrive on-line and are needed to be connected immediately. We give a simple O(log{sup 2} n) competitive deterministic on-line algorithm. The previous best online algorithm (by Westbrook and Yan) was O({radical}n log n) competitive. We also consider the network connectivity leasing problem which is a generalization of the GSP. Here edges of the graph can be either bought or leased for different costs. We provide simple randomized O(log{sup 2} n) competitive algorithm based on the on-line generalized Steiner problem result.

  18. The war against on-line piracy

    Harris, Julian


    A summary by Julian Harris, Deputy General Editor Amicus Curiae, of US attempts to control what it identifies as rogue Internet sites engaged in on-line piracy and opposition to such legislative moves.

  19. On-Line Acquisitions by LOLITA

    Frances G. Spigai


    Full Text Available The on-line acquisition program (LOLITA in use at the Oregon State University Library is described in terms of development costs, equipment requirements, and overall design philosophy. In particular, the record format and content of records in the on-order file, and the on-line processing of these records (input, search, correction, output using a cathode ray tube display terminal are detailed.

  20. ACSEPP On-Line Electronic Payment Protocol

    WANG Shao-bin; ZHU Xian; HONG Fan


    With analyzing the existing on-line electronic payment protocols, this paper presents a new on-line electronic payment protocol named ACSEPP: Anonymous, Convenient and Secure Electronic Payment Protocol.Its aim is to design a practical electronic payment protocol which is both secure and convenient.Without using PKI_CA frame, it realized the anonymity of consumer and merchant, the convenient of handling, the low cost of maintenance and the security.

  1. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)


    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  2. Formación on line

    O. Grau-Perejoan

    Full Text Available La formación on line es una modalidad de enseñanza a distancia basada en las nuevas tecnologías. En este artículo se pretende hacer una introducción a base de describir a grandes rasgos sus características principales: asincronía, no presencialidad, comunicación escrita, función del profesor on line, así como los retos, los riesgos, las ventajas y los inconvenientes que plantea. Se exponen las diferencias entre la formación on line y la formación presencial, de manera que los docentes puedan adaptar de la mejor manera posible sus propuestas formativas a la modalidad on line. Se introduce el importantísimo papel de la planificación y de la fase de diseño y, finalmente, se repasan conceptos útiles para comprender mejor el mundo de la formación on line como son los conceptos entorno virtual de aprendizaje (EVA o Blended Learning (B-Learning.

  3. GnRH激动剂主动免疫对GnRHR在腺垂体与子宫表达及分布的作用研究%GnRH agonist active immunization differentially influences the expression and distribution of GnRHR in pituitary and uterus of ewes

    魏锁成; 巩转娣; 董江陵; 韦敏; 谢坤; 张杰; 王江川


    To study the expression of GnRH receptor (GnRHR) mRNA in the pituitary and GnRHR proteins in the uteri, as well as the GnRHR distribution in the uteri of ewes which were actively immunized with alarelin antigen, and also to explore the mechanisms of GnRH agonists regulating reproductive function in ewes. Twenty-eight ewes were randomly divided into four groups (n=l). The eves in the experimental group Ⅰ (EG- Ⅰ ), experimental group Ⅱ (EG- Ⅱ) and experimental group Ⅲ (EG-Ⅲ) were injected subcutaneously with 200 μg, 300 μg and 400 μg GnRH agonist (Alarelin) antigen twice (on day 0 and 14), respectively. Animals in the control group (CG) were injected 2.0 ml solvent subcutaneously twice (on day 0 and 14). The pituitary and uterine horns in each ewe were collected aseptically on day 70. The fluorescent quantitative PCR (FQ-PCR) was implemented to detect the expression of GnRHR mRNA in the pituitary. The Western blotting was also performed to measure the GnRHR proteins in the uteri. Immuno-histoehemistry SP (Streptomyces avidin-peroxidase) method and image analysis were used to locate and analyze the GnRHR distribution in the uteri. The results showed that the expression values of GnRHR mRNA in the EG- Ⅰ , EG- Ⅱ and EG- Ⅲ groups decreased, and were lower than that in CG, with the minimum in EG- Ⅲ (P<0.01). Compared to CG, the values of GnRHR proteins in EG- Ⅰ , EG-Ⅱ and EG-Ⅲ reduced by 3.46%, 4.90% and 24.78% (P < 0.05), respectively. GnRHR distributed mainly in cytoplasms and nucleus of the uterine endometrial cells and glandular epithelial cells. The gray scales of EG-Ⅲ were lower lhan that of CG (P<0.05). In conclusion, alarelin antigen immunization could suppress the expression of GnRHR mRNA in the pituitary of the ewes, and inhibit the expression GnRHR proteins in uteri. GnRHR localize mainly in the nucleus and cytoplasm of the endometrial epithelial cells and glandular epithelial cells of the uteri. GnRH agonist immunity could

  4. On line routing per mobile phone

    Bieding, Thomas; Görtz, Simon; Klose, Andreas


    . Additionally it is of utmost importance that the employed communication system is suitable integrated with the firm’s enterprise application system and business processes. On basis of a case study, we describe in this paper a system that is cheap and easy to implement due to the use of simple mobile phones......On-line routing is concerned with building vehicle routes in an ongoing fashion in such a way that customer requests arriving dynamically in time are efficiently and effectively served. An indispensable prerequisite for applying on-line routing methods is mobile communication technology...

  5. Educational On-Line Gaming Propensity

    Sudzina, Frantisek; Razmerita, Liana; Kirchner, Kathrin


    Educational on-line games are promising for new generations of students who are grown up digital. Th e new generations of students are technology savvy and spend lots of time on the web and on social networks. Based on an exploratory study, this article investigates the factors that infl uence...... students’ willingness to participate in serious games for teaching/learning. Th is study investigates the relationship between students’ behavior on Facebook, Facebook games, and their attitude toward educational on-line games. Th e results of the study reveal that the early adopters of educational games...

  6. Optimal design of passive containment cooling system for innovative PWR

    Huiun Ha


    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  7. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Klaus Umminger


    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  8. Aqueous Nanofluid as a Two-Phase Coolant for PWR

    Pavel N. Alekseev


    Full Text Available Density fluctuations in liquid water consist of two topological kinds of instant molecular clusters. The dense ones have helical hydrogen bonds and the nondense ones are tetrahedral clusters with ice-like hydrogen bonds of water molecules. Helical ordering of protons in the dense water clusters can participate in coherent vibrations. The ramified interface of such incompatible structural elements induces clustering impurities in any aqueous solution. These additives can enhance a heat transfer of water as a two-phase coolant for PWR due to natural forming of nanoparticles with a thermal conductivity higher than water. The aqueous nanofluid as a new condensed matter has a great potential for cooling applications. It is a mixture of liquid water and dispersed phase of extremely fine quasi-solid particles usually less than 50 nm in size with the high thermal conductivity. An alternative approach is the formation of gaseous (oxygen or hydrogen nanoparticles in density fluctuations of water. It is possible to obtain stable nanobubbles that can considerably exceed the molecular solubility of oxygen (hydrogen in water. Such a nanofluid can convert the liquid water in the nonstoichiometric state and change its reduction-oxidation (RedOx potential similarly to adding oxidants (or antioxidants for applying 2D water chemistry to aqueous coolant.

  9. Mitsubishi PWR nuclear fuel with advanced design features

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)


    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  10. PWR safety/relief valve blowdown analysis experience

    Lee, M.Z.; Chou, L.Y.; Yang, S.H. (Gilbert/Commonwealth Engineers and Consultants, Reading, PA (USA). Speciality Engineering Dept.)


    The paper describes the difficulties encountered in analyzing a PWR primary loop pressurizer safety relief valve and power operated relief valve discharge system, as well as their resolution. The experience is based on the use of RELAP5/MOD1 and TPIPE computer programs as the tools for fluid transient analysis and piping dynamic analysis, respectively. General approaches for generating forcing functions from thermal fluid analysis solution to be used in the dynamic analysis of piping are reviewed. The paper demonstrates that the 'acceleration or wave force' method may have numerical difficulties leading to unrealistic, large amplitude, highly oscillatory forcing functions in the vicinity of severe flow area discontinuities or choking junctions when low temperature loop seal water is discharged. To avoid this problem, an alternate computational method based on the direct force method may be used. The simplicity and superiority in numerical stability of the forcing function computation method as well as its drawbacks are discussed. Additionally, RELAP modeling for piping, valve, reducer, and sparger is discussed. The effects of loop seal temperature on SRV and PORV discharge line blowdown forces, pressure and temperature distributions are examined. Finally, the effects of including support stiffness and support eccentricity in piping analysis models, method and modeling relief tank connections, minimization of tank nozzle loads, use of damping factors, and selection of solution time steps are discussed.




    Full Text Available This paper outlines the steps and technologies used in developing an on-line application server with many desktop clients, and with high power processing for a wide range of input documents to obtain searchable documents on the highest portability standards, PDF and PDF /A.

  12. SCC crack growth rate of cold-worked austenitic stainless steels in PWR primary water conditions

    Guerre, C.; Raquet, O.; Herms, E. [Commissariat a l' Energie Atomique (CEA), DEN/DPC/SCCME/LECA, Gif-sur-Yvette Cedex (France); Marie, S. [Commissariat a l' Energie Atomique (CEA), DEN/DM2S/SEMT/LISN, Gif-sur-Yvette Cedex (France); Le Calvar, M. [Inst. for Radiological Protection and Nuclear Safety (IRSN), DSR/SAMS, Fontenay-aux-Roses Cedex (France)


    Stress corrosion cracking (SCC) of stainless steels (SS) is a significant cause of failure in the pressurized water reactors (PWR). Most of the reported case history failures of SS in PWR can be attributed to pollutants (chloride, sulphate) and / or locally oxygenated environments, even to sensitisation of the SS. However, some failures have been attributed to heavy cold work (CW) of SS. In laboratory tests, SCC initiation of cold-worked SS has been obtained using slow strain rate tests (SSRT) in nominal PWR environment. This paper describes constant load and cyclic crack growth rate (CGR) tests on cold-worked SS, on CT specimens. 304L and 316L have been tested with a CW up to 60 %. CW 316L is more prone to cracking than 304L. Over 30 % of CW, 316L is susceptible to crack propagation under constant load. CW is the main controlling parameter for cracking. (author))


    Amir Hamzah


    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  14. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Xiangjie Liu


    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  15. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)


    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  16. Analyses of PWR boron dilution consequences with the Arrotta code

    Johanson, E.; Cheng, H.W.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety


    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced.

  17. Continuous firefly algorithm applied to PWR core pattern enhancement

    Poursalehi, N., E-mail: [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)


    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

  18. French nuclear plants PWR vessel integrity assessment and life management

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)


    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  19. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.


    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  20. Educational On-Line Gaming Propensity

    Sudzina, Frantisek; Razmerita, Liana; Kirchner, Kathrin


    Educational on-line games are promising for new generations of students who are grown up digital. Th e new generations of students are technology savvy and spend lots of time on the web and on social networks. Based on an exploratory study, this article investigates the factors that infl uence...... students’ willingness to participate in serious games for teaching/learning. Th is study investigates the relationship between students’ behavior on Facebook, Facebook games, and their attitude toward educational on-line games. Th e results of the study reveal that the early adopters of educational games...... are likely to be students, who are young, have only a few Facebook connections, who currently play Facebook game(s). Furthermore, the study emphasizes that there may be differences between students coming from various countries....

  1. New Trends in on-line Marketing

    Palkovič, Lukáš


    This bachelor thesis deals with new trend of internet marketing, it focuses especially on viral marketing. The theoretical part charasterizes the process of viral campaigns, furthermore deals with the components and aspects of on-line environment. Another separated chapter presents social networks, their place in viral marketing and at last but not least the viral video making process. The practical part contains different analyses of specific viral campaigns. The next and equally the last pa...

  2. Connecting to On-line Data

    Eichhorn, G.; Astrophysics Datacenter Executive Committee (ADEC)


    The Astrophysics Datacenter Executive Committee (ADEC) is coordinating the development of a system to facilitate the linking to on-line data. This system has three components: 1. Unique dataset identifiers. 2. A verification system for identifiers. 3. Permanent links to on-line data sets. 1. The ADEC has agreed on a naming scheme for data sets that allows for the unique identification of any data set. The ADEC data centers will clearly mark their data with these identifiers to allow the generation of links to these data. 2. Each data center has a utility that can check whether a data set identifier is a valid identifier at that center. A central verifier allows third parties access to these individual verifiers through a single portal. 3. The central verifier also provides permanent links to data sets through a central link forwarding system. This makes it possible to move data sets between data centers while maintaining the permanent links. The ADEC plans to first use this system to implement the linking from the literature to on-line data in a collaboration with the AAS and the University of Chicago Press for the AAS journals.

  3. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.


    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  4. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    Maucec, M; Glumac, B


    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  5. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    Maucec, M; Glumac, B


    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  6. Identification of dose-reduction techniques for BWR and PWR repetitive high-dose jobs

    Dionne, B.J.; Baum, J.W.


    As a result of concern about the apparent increase in collective radiation dose to workers at nuclear power plants, this project will provide information to industry in preplanning for radiation protection during maintenance operations. This study identifies Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) repetitive jobs, and respective collective dose trends and dose reduction techniques. 3 references, 2 tables. (ACR)

  7. Assessment of void swelling in austenitic stainless steel PWR core internals.

    Chung, H. M.; Energy Technology


    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling

  8. Total on-line purchasing system (TOPS)

    Collins, N.


    The Information Management Division (IMD) at LLNL is developing a new purchasing system for the Procurement Department. The first major development of this new system is called, {open_quotes}Total On-Line Purchasing System{close_quotes} (TOPS). TOPS will help speed up the requisitioning process by having requisitions electronically entered by requesters and electronically sent to buyers to be put on Purchase Orders. The new purchasing system will use Electronic Commerce (EC)/Electronic Data Interchange (EDI), to help increase transaction flows for shipping notices, RFQs, Quotes, Purchase Orders, and Invoices. ANSI X.12 is the EDI standard that this new EC will use.

  9. On-line and Mobil Learning Activities

    Ackerman, S. A.; Whittaker, T. M.; Jasmin, T.; Mooney, M. E.


    Introductory college-level science courses for non-majors are critical gateways to imparting not only discipline-specific information, but also the basics of the scientific method and how science influences society. They are also indispensable for student success to degree. On-line, web-based homework (whether on computers or mobile devices) is a rapidly growing use of the Internet and is becoming a major component of instruction in science, replacing delayed feedback from a few major exams. Web delivery and grading of traditional textbook-type questions is equally effective as having students write them out for hand grading, as measured by student performance on conceptual and problem solving exams. During this presentation we will demonstrate some of the interactive on-line activities used to teach concepts and how scientists approach problem solving, and how these activities have impacted student learning. Evaluation of the activities, including formative and summative, will be discussed and provide evidence that these interactive activities significantly enhance understanding of introductory meteorological concepts in a college-level science course. More advanced interactive activities are also used in our courses for department majors, some of these will be discussed and demonstrated. Bring your mobile devices to play along! Here is an example on teaching contouring:

  10. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.


    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  11. 基于CFD计算的核电厂半管水位运行工况余排接管入口涡流吸气效应研究%CFD simulation on Vortex and air-ingestion phenomenon at the junction of residual heat removal system (RHR) during mid-loop operation at nuclear power plant

    沈云海; 赵禹; 张玉龙; 赖建永; 王保平


    核电厂半管水位运行工况期间,余热排出系统与主管道连接的管道入口可能发生涡流吸气现象导致气体进入余排管道,影响余排泵的效率甚至造成余排泵气蚀损坏,进而导致余热导出能力失效。本文基于成熟的核电厂余排管道入口结构设计,采用CFD的方法对不同形式的余排管道入口流动情况进行模拟仿真计算分析,并对半管水位、余排流量等关键参数进行了对比分析,一定程度上为余排接管入口防涡流吸气的设计提供了理论指导。%Vortex and air-ingestion phenomenon may happen at junction of residual heat removal system (RHR) and primary pipe during mid-loop operation at nuclear power plant. Vortex and air-ingestion phenomenon leading to gas go inside RHR pipe can influence the efficiency of RHR pump, even damage the pump by cavitation and result in ability of removing heat of reactor lose finally. CFD were used to simulate fluid field at the junction, and analyse the parameter of fluid field such as level of mid-loop and flow rate. This article is base on mature design of structural of RHR junction, and it can help to design of preventing Air-ingestion phenomenon.

  12. The Power-weakness Ratios (PWR as a Journal Indicator: Testing the “Tournaments” Metaphor in Citation Impact Studies

    Loet Leydesdorff


    Full Text Available Purpose: Ramanujacharyulu developed the Power-weakness Ratio (PWR for scoring tournaments. The PWR algorithm has been advocated (and used for measuring the impact of journals. We show how such a newly proposed indicator can empirically be tested. Design/methodology/approach: PWR values can be found by recursively multiplying the citation matrix by itself until convergence is reached in both the cited and citing dimensions; the quotient of these two values is defined as PWR. We study the effectiveness of PWR using journal ecosystems drawn from the Library and Information Science (LIS set of the Web of Science (83 journals as an example. Pajek is used to compute PWRs for the full set, and Excel for the computation in the case of the two smaller sub-graphs: (1 JASIST+ the seven journals that cite JASIST more than 100 times in 2012; and (2 MIS Quart+ the nine journals citing this journal to the same extent. Findings: A test using the set of 83 journals converged, but did not provide interpretable results. Further decomposition of this set into homogeneous sub-graphs shows that—like most other journal indicators—PWR can perhaps be used within homogeneous sets, but not across citation communities. We conclude that PWR does not work as a journal impact indicator; journal impact, for example, is not a tournament. Research limitations: Journals that are not represented on the “citing” dimension of the matrix—for example, because they no longer appear, but are still registered as “cited” (e.g. ARIST—distort the PWR ranking because of zeros or very low values in the denominator. Practical implications: The association of “cited” with “power” and “citing” with “weakness” can be considered as a metaphor. In our opinion, referencing is an actor category and can be Metaphor in Citation Impact Studies in terms of behavior, whereas “citedness” is a property of a document with an expected dynamics very different from that of

  13. Radiative heat transfer modelling in a PWR severe accident sequence

    Magali Zabiego; Florian Fichot [Institut de Radioprotection et de Surete Nucleaire - BP 3 - 13115 Saint-paul-Lez-Durance (France); Pablo Rubiolo [Westinghouse Science and Technology - 1344 Beulah Road - Pittsburgh - PA 15235 (United States)


    a debris bed. In particular, an expression of the conductivity was established in cells in which remaining cylinders and debris particles coexist. In the present document, after a recall of the main lines of the modelling, an application to a reactor sequence is proposed. A severe accident transient with core degradation is simulated. The radiative transfer model is shown to behave properly and to smoothly calculate the transitions between the successive core configurations. A comparison with the more classical Hottel method shows that the present model gives a better prediction of the degradation progression owing to a more accurate estimate of the radial heat transfers. References: [1] M. Zabiego et al., ICARE/CATHARE V1: application to a PWR 900 MWe severe accident sequence, SARJ, Tokyo, 1999; [2] M. Zabiego, F. Fichot, P. Rubiolo Transfert radiatif lors d'une sequence accidentelle dans un coeur de Reacteur a Eau sous Pression, Congres Francais de Thermique, SFT 2004, Presqu'ile de Giens, 25-28 mai 2004. (authors)

  14. Aprender a innovar: una experiencia on line



    Full Text Available La creatividad y la innovación se han convertido en recursos clave en la denominada sociedad del conocimiento, que bien podría ser también llamada sociedad de la innovación. Pero innovar es una actividad compleja, que integra la aplicación de múltiples capacidades, el pensamiento divergente y convergente, la gestión de equipos humanos, la comunicación. Ahora bien, a innovar se puede, y se debe, aprender. Aprender a innovar es un reto y también una obligación para el conjunto del sistema educativo en todos sus niveles. Partiendo de estas consideraciones este trabajo expone una experiencia de aprendizaje de la creatividad y de la innovación a través de un curso totalmente on line basado en la plataforma MOODLE, en el marco del Programa de Formación Permanente de la Universidad de Cádiz. Se presenta un modelo del proceso de innovación, denominado CREALAB, de elaboración propia. Este modelo se ha utilizado como base del proceso de aprendizaje de la creatividad y de la innovación y en el diseño del curso, está organizado en torno a actividades y tiene un carácter iterativo y realimentado. Se presentan además el conjunto del diseño metodológico y los resultados obtenidos en las dos ediciones celebradas hasta el momento. El diseño del curso totalmente on line y los resultados alcanzados permiten estimar un alto potencial de aplicación, tanto a nivel personal como a nivel organizacional.

  15. On-line chemical composition analyzer development

    Roberts, M.J.; Garrison, A.A.; Muly, E.C.; Moore, C.F.


    The energy consumed in distillation processes in the United States represents nearly three percent of the total national energy consumption. If effective control of distillation columns can be accomplished, it has been estimated that it would result in a reduction in the national energy consumption of 0.3%. Real-time control based on mixture composition could achieve these savings. However, the major distillation processes represent diverse applications and at present there does not exist a proven on-line chemical composition sensor technology which can be used to control these diverse processes in real-time. This report presents a summary of the findings of the second phase of a three phase effort undertaken to develop an on-line real-time measurement and control system utilizing Raman spectroscopy. A prototype instrument system has been constructed utilizing a Perkin Elmer 1700 Spectrometer, a diode pumped YAG laser, two three axis positioning systems, a process sample cell land a personal computer. This system has been successfully tested using industrially supplied process samples to establish its performance. Also, continued application development was undertaken during this Phase of the program using both the spontaneous Raman and Surface-enhanced Raman modes of operation. The study was performed for the US Department of Energy, Office of Industrial Technologies, whose mission is to conduct cost-shared R D for new high-risk, high-payoff industrial energy conservation technologies. Although this document contains references to individual manufacturers and their products, the opinions expressed on the products reported do not necessarily reflect the position of the Department of Energy.

  16. Influence of visualization on consumption during on-line shopping

    Hictaler, Urška


    This diploma work studies the influence of visualization on consumption during on-line shopping. The first part of the thesis starts with key areas of visualization, consumption and on-line shopping. Visualization, areas of use, human perception and ways of product presentation in on-line shops are defined discussed first. Next, consumption, consumers and factors that influence their decisions and satisfaction are defined. The last topic in the first part of the thesis discusses on-line shopp...

  17. On-Line Learning: One Way to Bring People Together

    Goff-Kfouri, Carol Ann


    The purpose of this study was to demonstrate the benefits of on-line learning for adult learners and to further demystify three common misconceptions concerning on-line learning: students certainly do receive support from their on-line professors, the professor is pro-active rather than passive, and students may be more motivated to learn than in…

  18. A Distributed System for Learning Programming On-Line

    Verdu, Elena; Regueras, Luisa M.; Verdu, Maria J.; Leal, Jose P.; de Castro, Juan P.; Queiros, Ricardo


    Several Web-based on-line judges or on-line programming trainers have been developed in order to allow students to train their programming skills. However, their pedagogical functionalities in the learning of programming have not been clearly defined. EduJudge is a project which aims to integrate the "UVA On-line Judge", an existing…

  19. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail:; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail:; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)


    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  20. Eddy current NDT: a suitable tool to measure oxide layer thickness in PWR fuel rods

    Alencar, Donizete A.; Silva Junior, Silverio F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil)], e-mail:, e-mail:; Vieira, Andre L.P.S. [Industrias Nucleares do Brasil (INB S.A.), Resende, RJ (Brazil). Fabrica de Combustivel Nuclear], e-mail:; Soares, Adolpho [Technotest Consultoria e Acessoria Ltda., Belo Horizonte, MG (Brazil)], e-mail:


    Eddy current is a nondestructive test (NDT) widely used in industry to support integrity analysis of components and equipment. In the nuclear area it is frequently applied to inspect tubes installed in tube exchangers, such as steam generators and condensers in PWR plants, as well as turbine blades. Adequately assisted by means of robotic devices, that inspection method has been pointed as a suitable tool to perform accurate oxide layer thickness measurements in PWR fuel rods. This paper shows some theoretical aspects and physical operating principles of the inspection method, as well as test probes construction details, and the calibration reference standards fabrication processes. Furthermore, some data, experimentally obtained at INB laboratories and other technical information obtained from TECNATOM S.A. are presented, showing the accuracy and efficacy of such NDT method. (author)

  1. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Herranz, L. E.; Lopez, C.


    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.


    Pande Made Udiyani


    Full Text Available ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe.  Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor.  Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Data meteorologi diambil setiap jam untuk jangka waktu satu tahun. Hasil perhitungan menunjukkan bahwa frekuensi kumulatif  untuk model input yang sama untuk Tapak pesisir Serang lebih tinggi dibandingkan dengan Semenanjung Muria. Untuk tapak yang sama, frekuensi kumulatif model input cyclic lebih tinggi dibandingkan model stratified. Model cyclic memberikan keleluasan dalam menentukan tingkat ketelitian perhitungan dan tidak membutuhkan data acuan dibandingkan dengan model stratified. Penggunaan model cyclic dan stratified melibatkan jumlah data yang besar dan pengulangan perhitungan  akan meningkatkan  ketelitian nilai-nilai statistika perhitungan. Kata kunci: dampak kecelakaan, PWR 1000-MWe,  probabilistik,  atmosferik, PC-Cosyma   ABSTRACT THE INFLUENCE OF ATMOSPHERIC CONDITIONS TO PROBABILISTIC CALCULATION OF IMPACT OF RADIOLOGY ACCIDENT ON PWR-1000MWe. The calculation of the radiological impact of the fission products releases due to potential accidents

  3. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)


    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  4. EPRI PWR Safety and Relief Value Test Program: safety and relief valve test report


    A safety and relief valve test program was conducted by EPRI for a group of participating PWR utilities to respond to the USNRC recommendations documented in NUREG 0578 Section 2.1.2, and as clarified in NUREG 0737 Item II.D.1.A. Seventeen safety and relief valves representative of those utilized in or planned for use in participating domestic PWR's were tested under the full range of selected test conditions. This report contains a listing of the selected test valves and the corresponding as tested test matrices, valve performance data and principal observations for the tested safety and relief valves. The information contained in this report may be used by the participating utilities in developing their response to the above mentioned USNRC recommendations.

  5. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.


    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  6. On-Line Impact Load Identification

    Krzysztof Sekuła


    Full Text Available The so-called Adaptive Impact Absorption (AIA is a research area of safety engineering devoted to problems of shock absorption in various unpredictable scenarios of collisions. It makes use of smart technologies (systems equipped with sensors, controllable dissipaters and specialised tools for signal processing. Examples of engineering applications for AIA systems are protective road barriers, automotive bumpers or adaptive landing gears. One of the most challenging problems for AIA systems is on-line identification of impact loads, which is crucial for introducing the optimum real-time strategy of adaptive impact absorption. This paper presents the concept of an impactometer and develops the methodology able to perform real-time impact load identification. Considered dynamic excitation is generated by a mass M1 impacting with initial velocity V0. An analytical formulation of the problem, supported with numerical simulations and experimental verifications is presented. Two identification algorithms based on measured response of the impacted structure are proposed and discussed. Finally, a concept of the AIA device utilizing the idea of impactometer is briefly presented.


    Rubén Faúndez


    Full Text Available Las aplicaciones de simulación tienden a ser cada vez más cercanas a usuarios e industrias. Sin embargo, muchas de ellas no poseen ni la capacidad ni el conocimiento como para desarrollar internamente sus modelos de simulación. Por este motivo, y como una forma de apoyar la toma de decisiones basándose en modelos de simulación, se presenta la plataforma SOL (Simulación On Line. La metodología completa de trabajo, así como la interacción entre SOL, Empresa y Asesor, son presentadas. Su base de datos, los niveles de usuarios, sus funcionalidades, y la creación automatizada de información grafica y visual, también son explicadas. En el caso de aplicación, el uso de SOL para apoyar la toma de decisiones en una operación de movimiento de material, permite a los tomadores de decisión acceder a análisis robustos basados en información extraída de los modelos de simulación. SOL, al almacenar información, funcionar vía web, generar análisis automatizados y crear visualizaciones, permite cumplir con las expectativas de los usuarios respecto a una solución integral en simulación.

  8. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    Cardoni, Jeffrey


    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  9. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  10. Anti -corrosion Effect of ETA on Materials in Secondary Loop of PWR


    In the world, over sixty percent of nuclear power plant have used advanced amunes ETA(Ethanolamine) as pH control agent in secondary loop of PWR. There are eighty percent of nuclear powerplants using ETA in USA. The corrosion of materials in steam generator (SG) tube and secondary looppower water reactor have been inhibited, the life of SG and the economics of the plant are increasedbecause of using ETA.

  11. Modeling and Simulation of Release of Radiation in Flow Blockage Accident for Two Loops PWR

    Khurram Mehboob; Cao Xinrong; Majid Ali


    In this study modeling and simulation of release of radiation form two loops PWR has been carried out for flow blockage accident. For this purpose, a MATLAB based program “Source Term Evaluator for Flow Blockage Accident” (STEFBA) has been developed, which uses the core inventory as its primary input. The TMI-2 reactor is considered as the reference plant for this study. For 1100 reactor operation days, the core inventory has been evaluated under the core design constrains at average reactor ...

  12. Chemical and radiochemical specifications - PWR power plants; Specifications chimiques et radiochimiques - Centrales REP

    Stutzmann, A. [Electricite de France (EDF), 93 - Saint-Denis (France)


    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  13. Proof test on thermal and hydraulic design reliability of Japanese PWR fuel assemblies

    Akiyama, Mamoru (Univ. of Tokyo (Japan)); Inoue, Akira (Tokyo Institute of Technology (Japan)); Miyazaki, Keiji (Osaka Univ. (Japan)); Abeta, Sadaaki (Mitsubishi, Tokyo (Japan)); Hori, Keiichi (Mitsubishi, Hyogo (Japan)); Mukasa, Tomio; Oishi, Masao; Aoki, Toshimasa; Makihara, Yoshiaki


    A series of departure from nucleate boiling (DNB) tests for pressurized water reactors (PWRs) was performed at the Nuclear Power Engineering Test Center. The objective was to prove the reliability of fuel assembly design by confirming the thermal margin of heat transfer. The present method for evaluating the DNB ratio in a Japanese 17 x 17 PWR core is adequate according to the newly obtained DNB test data.

  14. EDF/CIDEN - ONECTRA: PWR decontamination; EDF/CIDEN - ONECTRA: assainissement REP

    Fayolle, P. [EDFICIDEN, 35-37, rue Louis Guerin - B.P. 21212, 69611 Villeurbanne Cedex (France); Orcel, H. [ONECTRA, ZA les Tomples BP45, 26701 Pierrelatte Cedex (France); Wertz, L. [ONECTRA, Le Britannia, Allee C, 20 Bd Eugene Deruelle, 69432 Lyon Cedex 03 (France)


    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  15. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)


    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  16. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    Naegeli, Robert Earl


    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  17. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)


    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  18. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.


    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

  19. Evaluation of the RELAP4/MOD6 thermal-hydraulic code. [PWR

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


    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.

  20. Robust on-line monitoring of biogas processes; Robusta maettekniker on-line foer optimerad biogasproduktion

    Nordberg, Aake; Hansson, Mikael; Kanerot, Mija; Krozer, Anatol; Loefving, Bjoern; Sahlin, Eskil


    Although demand for biomethane in Sweden is higher than ever, many Swedish codigestion plants are presently operated below their designed capacity. Efforts must be taken to increase the loading rate and guarantee stable operation and high availability of the plants. There are currently no commercial systems for on-line monitoring, and due to the characteristics of the material, including corrosion and tearing, robust applications have to be developed. The objective of this project was to identify and study different monitoring technologies with potential for on-line monitoring of both substrate mixtures and anaerobic digester content. Based on the prerequisites and demands at Boraas Energi och Miljoe AB's (BEMAB, the municipal energy and waste utility in the city of Boraas, Sweden) biogas plant, the extent of the problems, measurement variables and possible ways of managing these issues have been identified and prioritized. The substrate mixtures in question have a high viscosity and are inhomogeneous with variation in composition, which calls for further homogenization, dilution and filtration to achieve high precision in the necessary analyses. Studies of using different mixers and mills showed that the particle size (800 mum) needed for on-line COD measurement could not be achieved. The problem of homogenization can be avoided if indirect measurement methods are used. Laboratory tests with NIR (near-infra red spectroscopy) showed that VS can be predicted (R2=0,78) in the interval of 2-9% VS. Furthermore, impedance can give a measurement of soluble components. However, impedance is not sensitive enough to give a good measurement of total TS. Microwave technology was installed at the production plant and showed a faster response to changes in TS than the existing TS-sensor. However, due to technical problems, the evaluation only could be done during a limited period of ten days. BEMAB will continue the measurements and evaluation of the instrument. The

  1. DOE-EPRI On-Line Monitoring Implementation Guidelines

    E. Davis, R. Bickford


    Industry and EPRI experience at several plants has shown on-line monitoring to be very effective in identifying out-of-calibration instrument channels or indications of equipment-degradation problems. The EPRI implementation project for on-line monitoring has demonstrated the feasability of on-line monitoring at several participating nuclear plants. The results have been very enouraging, and substantial progress is anticipated in the coming years.

  2. Poss On-line (Personalisation of Self-Service Solutions across On-line platforms)

    Nielsen, Janni; Nielsen, Lene; Jespersen, Mikkel


    The project on Personalisation of Self-service Solutions across On-line Platforms (POSS ON-LINE) focuses on users, clients, and self-service solutions. It is based on the understanding that clients and users are different and have different goals, and that self-service takes place in different co...... at the process of development. However, we lack methods to predict user behaviour without having to deal with huge amounts of data and data from both quantitative data as well as life world observations are required....... the client and the user. The system gathers data about the user, which enables the client to push information to the user. Personalisation enables graphic user interface design that is personalised and relevant to the individual user and invites the user to get access to information with less strain....... Personalisation of self-service solutions is promising and IT companies are experiencing an increase in the clients' demands. At the same time the development of solutions moves within a shorter and shorter time span. Hence the process of innovations is paced and there is an increasing need of new ways of looking...

  3. Study of the distribution of hydrogen in a PWR containment with CFD codes; Estudio de la distribucion de hidrogeno en una contencion PWR con codigos CFD

    Jimenez, G.; Martinez, R. M.; Fernandez, K.; Morato, D. J.; Bocanegra Melian, R.; Mena, L.; Queral, C.


    During the development of a severe accident in a PWR reactor can be generated, large quantities of hydrogen by the oxidation of metals present in the nucleus, mainly the zirconium fuel pods. This hydrogen, along with steam and other gases, can be released to the atmosphere of contention by a leak or break in the primary circuit and achieving conditions in which combustion may occur. Combustion causes thermal and pressure loads that can damage the security systems and the integrity of the containment building, last barrier of confinement of radioactive materials. The main condition that defines the characteristics of the combustion is the concentration of species, detailed knowledge of the distribution of hydrogen is very important to correctly predict the possible damage to the containment in the event that there is combustion. (Author)

  4. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Alves, Antonio Carlos Pinto Dias


    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  5. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Lapa, Nelbia da Silva


    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

  6. Valve inlet fluid conditions for pressurizer safety and relief valves in Westinghouse-designed plants. Final report. [PWR

    Meliksetian, A.; Sklencar, A.M.


    The overpressure transients for Westinghouse-designed NSSSs are reviewed to determine the fluid conditions at the inlet to the PORV and safety valves. The transients considered are: licensing (FSAR) transients; extended operation of high pressure safety injection system; and cold overpressurization. The results of this review, presented in the form of tables and graphs, define the range of fluid conditions expected at the inlet to pressurized safety and power-operated relief valves utilized in Westinghouse-designed PWR units. These results will provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI/PWR Safety and Relief Valve Test Program indeed envelop those expected in their units.


    Andi Sofrany Ekariansyah


    Full Text Available AP1000 adalah reaktor daya PWR maju dengan daya listrik 1154 MW yang didesain berdasarkan kinerja teruji dari desain PWR lain oleh Westinghouse. Untuk mempersiapkan peran Pusat Teknologi Reaktor dan Keselamatan Nuklir sebagai suatu Technical Support Organization (TSO dalam hal verifikasi keselamatan, telah dilakukan kegiatan verifikasi keselamatan untuk AP1000 yang dimulai dengan verifikasi kecelakaan kegagalan pendingin sekunder. Kegiatan dimulai dengan pemodelan fitur keselamatan teknis yaitu sistem pendinginan teras pasif yang terdiri dari sistem Passive Residual Heat Removal (PRHR, tangki core makeup tank (CMT, dan tangki In-containment Refueling Water Storage Tank (IRWST. Kecelakaan kegagalan pendingin sekunder yang dipilih adalah hilangnya aliran air umpan ke salah satu pembangkit uap yang disimulasikan menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Tujuan analisis adalah untuk memperoleh sekuensi perubahan parameter termohidraulika reaktor akibat kecelakaan dimana hasil analisis yang diperoleh divalidasi dan dibandingkan dengan hasil analisis menggunakan program perhitungan LOFTRAN di dalam dokumen desain keselamatan AP1000. Hasil verifikasi menunjukkan bahwa kejadian hilangnya suplai air umpan tidak berdampak pada kerusakan teras, sistem pendingin reaktor, maupun sistem sekunder. Penukar kalor PRHR telah terverifikasi kemampuannya dalam membuang kalor peluruhan teras setelah trip reaktor. Hasil validasi dengan dokumen pembanding menunjukkan kesesuaian pada sebagian besar parameter termohidraulika. Secara umum, model PWR maju yang dilengkapi dengan sistem pendinginan teras ciri pasif yang telah dikembangkan tetap selamat ketika terjadi kecelakaan kehilangan aliran pendingin sekunder. Kata kunci: Verifikasi, hilangnya aliran air umpan, AP1000   AP1000 is a PWR power reactor with 1154 MW of electrical power that is designed based on the proven performance of the other Westinghouse PWR designs. To prepare the role of Center for

  8. PWR circuit contamination assessment tool. Use of OSCAR code for engineering studies at EDF

    Benfarah Moez


    Full Text Available Normal operation of PWR generates corrosion and wear products in the primary circuit which are activated in the core and constitute the major source of the radiation field. In addition, cases of fuel failure and alpha emitter dissemination in the coolant system could represent a significant radiological risk. Radiation field and alpha risks are the main constraints to carry out maintenance and to handle effluents. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products and actinides and to carry out the appropriate measurements in PWR circuits and loop experiments. As a matter of fact, it is more than necessary to develop and use a reactor contamination assessment code in order to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has actually been developed and used for this aim. It is presented in this paper, as well as its use in the engineering studies at EDF. To begin with, the code structure is described, including the physical, chemical and transport phenomena considered for the simulation of the mechanisms regarding PWR contamination. Then, the use of OSCAR is illustrated with two examples from our engineering studies. The first example of OSCAR engineering studies is linked to the behavior of the activated corrosion products. The selected example carefully explores the impact of the restart conditions following a reactor mid-cycle shutdown on circuit contamination. The second example of OSCAR use concerns fission products and disseminated fissile material behavior in the primary coolant. This example is a parametric study of the correlation between the quantity of disseminated fuel and the variation of Iodine 134 in the primary coolant.

  9. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.


    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  10. Fatigue Life of Stainless Steel in PWR Environments with Strain Holding

    Kim, Taesoon; Kim, Kyuhyung [KHNP CRI, Daejeon (Korea, Republic of); Seo, Myeonggyu; Jang, Changheui [KAIST, Daejeon (Korea, Republic of)


    Many components and structures of nuclear power plants are exposed to the water chemistry conditions during the operation. Recently, as design life of nuclear power plant is expanded over 60 years, the environmentally assisted fatigue (EAF) due to these water chemistry conditions has been considered as one of the important damage mechanisms of the safety class 1 components. Therefore, many studies to evaluate the effect of light water reactor (LWR) coolant environments on fatigue life of materials have been conducted. Many EAF test results including Argonne National Laboratory’s consistently indicated the substantial reduction of fatigue life in the light water reactor environments. However, there is a discrepancy between laboratory test data and plant operating experience regarding the effects of environment on fatigue: while laboratory test data suggest huge accumulation of fatigue damage, very limited experience of cracking caused by the low cycle fatigue in light water reactor. These hold-time effect tests are preformed to characterize the effects of strain holding on the fatigue life of austenitic stainless steels in PWR environments in comparison with the existing fixed strain rate results. Low cycle fatigue life tests were conducted for the type 316 stainless steel in 310℃ air and PWR environments with triangular strain. In agreement with the previous reports, the LCF life was reduced in PWR environments. Also for the slower strain rate, the reduction of LCF life was greater than the faster strain rate. The LCF test conditions for the hold-time effects were determined by the references and consideration of actual plant transient. To simulate the heat-up and cooldown transient, sub-peak strain holding during the down-hill of strain amplitude was chosen instead of peak strain holding which used in the previous researches.

  11. Effect of co-free valve on activity reduction in PWR

    Bahn, C.B.; Han, B.C.; Bum, J.S.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Lee, C.B. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)


    Radioactive nuclei, such as {sup 68}Co and {sup 60}Co, deposited on out-of-core surfaces in a pressurized water reactor (PWR) primary coolant system, are major sources of occupational radiation exposure to plant maintenance personnel and act as costly impediment to prompt and effective repairs. Valve hardfacing alloys exposed to primary coolant are considered as one of the main Co sources. To evaluate the Co-free valve, such as NOREM 02 and Deloro 50, the candidates for the alternative to Stellite 6, in a simulated PWR primary condition, SNU corrosion test loop (SCOTL) was constructed. For gate valves hard-faced with made of NOREM 02 and Deloro 50 hot cycling tests were conducted for up to 2,000 on-off cycles with cold leak tests at 1,000 cycle interval. It was observed that the leak rate of NOREM 02 (Fe-base) did not satisfy the nuclear grade valve leak criteria. After 1000 cycles test, while there was no leakage in case of Deloro 50 (Ni-base). Also, Deloro 50 showed no leakage after 2000 cycles. To estimate the activity reduction effect, we modified CRUDSIM-MIT which modeled the effects of coolant chemistry on the crud transport and activity buildup in the primary system of PWR. In the new code, crud evaluation and assessment (CREAT), {sup 60}Co activity buildup prediction includes 1) Co-base valve replacement effect, 2) Co-base valve maintenance effect, and 3) control rod drive mechanism (CRDM) and main coolant pump (MCP) shaft contribution. CREAT predicted that the main contributor of Co activity buildup was the corrosion-induced release of Co from the steam generator (SG) tubing. With new SG's tubed with alloy 690, Korean Next Generation Reactor (APR-1400) is expected to have about 64% lower Co activity on SG surface. The use of all Co-free valves is expected to cut additional 8% of activity which is only marginal. (authors)

  12. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  13. Evolution of reactor monitoring and protection systems for PWR; Evolution des systemes de surveillance et de protection des REP

    Chaloin, B. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France); Mourlevat, J.L. [FRAMATOME ANP, 92 - Paris-La-Defence (France)


    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  14. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement


    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  15. Application of LBB to high energy piping systems in operating PWR

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)


    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  16. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor


    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  17. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  18. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT

    Uwe Imke


    Full Text Available SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT. The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  19. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code. [PWR

    Ramsthaler, J. A.; Lime, J. F.; Sahota, M. S.


    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A.

  20. The continued development of the MFM suite and its practical application on a PWR system

    Thunem, Harald P-J; Zhang, Xinxin


    This paper reports on the results from the practical application of the Shape Shifter framework on the continued development of a graphical editing suite, the MFM Suite, for MFM and process model design and analysis. The primary use of the MFM Suite is diagnosis and prognosis of anomalies...... in physical processes. One of the Halden Reactor Project’s advanced NPP simulators based on a PWR is used to demonstrate the applicability of the suite in realistic situations. The paper presents a summary and suggests some plans for future research and development....

  1. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)


    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  2. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering


    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  3. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)


    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  4. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL


    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.


    ZHANG Yuzhong; WANG Shouyang; Bo Chen; ZHANG Shuxia


    We address the problem of preemptively schedule on-line jobs on arbitrary muniformly related machines with the objective of minimizing the schedule length. We provide the first on-line algorithm for this general problem, and show that the algorithm being the speeds of the m machines.

  6. From Off-line to On-line Handwriting Recognition

    Lallican, P.; Viard-Gaudin, C.; Knerr, S.


    On-line handwriting includes more information on time order of the writing signal and on the dynamics of the writing process than off-line handwriting. Therefore, on-line recognition systems achieve higher recognition rates. This can be concluded from results reported in the literature, and has been

  7. On-Line and Off-Line Assessment of Metacognition

    Saraç, Seda; Karakelle, Sema


    The study investigates the interrelationships between different on-line and off-line measures for assessing metacognition. The participants were 47 fifth grade elementary students. Metacognition was assessed through two off-line and two on-line measures. The off-line measures consisted of a teacher rating scale and a self-report questionnaire. The…

  8. On-line characterization of a hybridoma cell culture process.

    Zhou, W; Hu, W S


    The on-line determination of the physiological state of a cell culture process requires reliable on-line measurements of various parameters and calculations of specific rates from these measurements. The cell concentration of a hybridoma culture was estimated on-line by measuring optical density (OD) with a laser turbidity probe. The oxygen uptake rate (OUR) was determined by monitoring dynamically dissolved oxygen concentration profiles and closing oxygen balances in the culture. The base addition for neutralizing lactate produced by cells was also monitored on-line via a balance. Using OD and OUR measurements, the specific growth and specific oxygen consumption rates were determined on-line. By combining predetermined stoichiometric relationships among oxygen and glucose consumption and lactate production, the specific glucose consumption and lactate production rates were also calculated on-line. Using these on-line measurements and calculations, the hybridoma culture process was characterized on-line by identifying the physiological states. They will also facilitate the implementation of nutrient feeding strategies for fed-batch and perfusion cultures. (c) 1994 John Wiley & Sons, Inc.

  9. On-line optimal control of greenhouse crop cultivation.

    Straten, van G.


    Thus far, optimal control has primarily been investigated for seasonal crop growth optimization. On-line aspects have received much less attention. The decomposition between long term strategies and on-line control, however, is not trivial. Appreciable losses occur when set-points generated by seaso

  10. On-line optimal control of greenhouse crop cultivation.

    Straten, van G.


    Thus far, optimal control has primarily been investigated for seasonal crop growth optimization. On-line aspects have received much less attention. The decomposition between long term strategies and on-line control, however, is not trivial. Appreciable losses occur when set-points generated by seaso

  11. Algorithms for semi on-line multiprocessor scheduling problems


    In the classical multiprocessor scheduling problems, it is assumed that the problems are considered in off-line or on-line environment. But in practice, problems are often not really off-line or on-line but somehow in between. This means that, with respect to the on-line problem, some further information about the tasks is available, which allows the improvement of the performance of the best possible algorithms. Problems of this class are called semi on-line ones. The authors studied two semi on-line multiprocessor scheduling problems, in which, the total processing time of all tasks is known in advance, or all processing times lie in a given interval. They proposed approximation algorithms for minimizing the makespan and analyzed their performance guarantee. The algorithms improve the known results for 3 or more processor cases in the literature.

  12. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Moreno Chamorro, P.; Gallego Diaz, C.


    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  13. Estimate of the speed of the refrigerant on a PWR: three way based on the analysis of noise; Estimacion de la volecidad del refrigerante en un PWR: tres vias basadas en el analisis de ruido

    Montalvo, C.; Ruiz, M.; Garcia Berrocal, A.


    The speed of the refrigerant is a key parameter in the monitoring of the operation a PWR. He know this value and be able to track on-site It allows an understanding of the State of the kernel with valuable information about the refrigerant, and thus behavior on heat exchange which takes place in the reactor. (Author)

  14. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    XU Mi


    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  15. Accelerated IGA/SCC testing of Alloy 600 in contaminated PWR environments

    Miglin, B.P.; Sarver, J.M. [Babcock & Wilcox R& D Division, Alliance, OH (United States); Aoki, K. [NFI, Osaka (Japan); Koch, D.W. [Babcock & Wilcox Nuclear Services, Lynchburg, VA (United States); Takamatsu, H. [Kansai Electric, Osaka (Japan)


    An accelerated corrosion test (360{degrees}C for 2000 hrs) was performed on C-ring specimens machined from one heat of Alloy 600 tubing in the mill-annealed condition. The specimens were exposed to secondary-side pressurized-water-reactor (PWR) solutions contaminated with lead, sulfur, silicon, and a combination of these contaminants. Where possible, MULTEQ calculations were performed to determine the chemical concentrations so that a constant elevated-temperature pH of 4.5 was achieved. This test was designed to examine the ability of these contaminants to cause intergranular attack and/or stress corrosion in stressed Alloy 600 tubing. The results from this test demonstrated that under the test conditions used, lead-contaminated PWR secondary water induces and propagates intergranular attack (IGA) and stress corrosion cracking (SCC) in Alloy 600. Attack was intergranular; the degree of attack did not vary in the liquid or vapor portions of the test environments. Although attack was more severe at higher stresses, significant attack was observed in samples stressed to the typical operating stress. Solutions of only sulfur and only silicon displayed no initiation or propagation of either IGA or SCC. However, the solution containing all three contaminants caused attack with identical morphology to that observed in the lead-contaminated solution.

  16. Thermal hydraulic investigations and optimization on the EVC system of a PWR by CFD simulation

    Xi, Mengmeng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Zhang, Dalin, E-mail: [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China); Tang, Mao [China Nuclear Power Design Engineering Co., Ltd., 518124 Shenzhen (China); Wang, Chenglong; Zheng, Meiyin; Qiu, Suizheng [Department of Nuclear Science and Technology, State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, 710049 Xi’an (China)


    Highlights: • This study constructs a full CFD model for the EVC system of a PWR. • The complex fluid and solid coupling is treated in the computation. • Primary characteristics of the velocity, pressure and temperature distributions in the EVC system are investigated. • The optimization of the EVC system with different inlet boundaries are performed. - Abstract: In order to optimize the design of Reactor Pit Ventilation (EVC) system in a Pressurized Water Reactor (PWR), it is necessary to study the characteristics of the velocity, pressure and temperature fields in the EVC system. A full computational fluid dynamics (CFD) model for the EVC system is constructed by a commercial CFD code, where the complex fluid and solid coupling is treated. The Shear Stress Transport (SST) model is adopted to perform the turbulence calculation. This paper numerically investigates the characteristics of the velocity, pressure and temperature distributions in the EVC system. In particular, the effects of inlet air parameters on the thermal hydraulic characteristics and the reactor pit structure are also discussed for the EVC system optimization. Simulations are carried out with different mesh sizes and boundary conditions for sensitivity analysis. The computational results are important references to optimize the design and verify the rationality of the EVC system.

  17. PWR Containment Shielding Calculations with SCALE6.1 Using Hybrid Deterministic-Stochastic Methodology

    Mario Matijević


    Full Text Available The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS were demonstrated when applied to a realistic deep penetration Monte Carlo (MC shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR parameters is based on deterministic transport theory (SN method providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence between SN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

  18. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)


    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  19. Characterization of Oxide Layer with Precipitates of HANA-6 Exposed in Simulated PWR Primary Water Environment

    Jang, Hun; Lim, Jea Young; Lee, Sung Yong; Kim, Yoon Ho; Mok, Yong Kyoon [KEPCO NF, Daejeon (Korea, Republic of)


    The delayed oxidation behaviors of β-Nb ppts and their amorphization behaviors in HANA-6 and other Zr-base alloys have been frequently reported. On the other hand, although Zr(Nb,Fe)2 ppts could be formed in the HANA-6 alloy due to Fe impurities contained in Zrsponge, the oxidation behavior of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy has not been fully understood. In this study, oxide characteristics of HANA-6 corroded in simulated PWR environment for 165 and 315 days were investigated. And, oxidation behaviors of Zr(Nb,Fe)2 ppts contained in HANA-6 alloy were investigated by TEM with EDS techniques. The superior corrosion property of HANA-6 has been confirmed through corrosion test in simulated PWR water for 387 days. By using TEM/EDS technique, the oxide characteristics with presence of β- Nb (or β-enriched), and ZrNbFe (possibly Zr(Nb,Fe){sub 2}) ppts have been characterized as follows. 1. Delayed oxidation behaviors of β-Nb and Zr(Nb,Fe){sub 2} ppts and their amorphization due to oxidation were observed from TEM/EDS analyses. 2. The oxide layers having crystallite and partially amorphous ppts were slightly increased with increasing corrosion test time from 165 days to 315 days. 3. In outer oxide layer, Fe in Zr(Nb,Fe){sub 2} ppt was depleted and dissolved to outer layer of ppt and bulk oxide layer.

  20. Experiment data report for semiscale Mod-1 Test S-06-5. (LOFT counterpart test). [PWR



    Recorded test data are presented for Test S-06-5 of the Semiscale Mod-1 LOFT counterpart test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-06-5 was conducted from initial conditions of 2272 psia and 536/sup 0/F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold legs of the intact and broken loops to simulate emergency core coolant injection in a PWR. The purpose of Test S-06-5 was to assess the influence of the break nozzle geometry on core thermal and system response and on the subcooled and low quality mass flow rates at the break locations.

  1. Vulnerability of a partially flooded PWR reactor cavity to a steam explosion

    Cizelj, Leon [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)]. E-mail:; Koncar, Bostjan [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia); Leskovar, Matjaz [' Jozef Stefan' Institute Jamova 39, SI 1000 Ljubljana (Slovenia)


    When the hot molten core comes into contact with the water in the reactor cavity a steam explosion may occur. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and later, during the expansion of the water vapour, to production of missiles that may endanger surrounding structures. The purpose of the performed analysis is to provide an estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion, and to make an assessment of the vulnerabilities of the typical PWR cavity structures to steam explosions. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by comprehensive and reasonably conservative analyses of two typical ex-vessel steam explosion cases differing in the steam explosion energy conversion ratio. In particular, the vulnerability of the surrounding reinforced concrete walls to damage has been sought in both cases.

  2. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)


    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  3. Construction and utilization of linear empirical core models for PWR in-core fuel management

    Okafor, K.C.


    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k{sub {infinity}} profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results.

  4. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)


    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  5. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Surian Pinem


    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  6. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    DeHart, M.D.; Hermann, O.W.


    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  7. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Freire, Luciano Ondir, E-mail:; Andrade, Delvonei Alves de, E-mail:


    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  8. Gas entrainment by one single French PWR spray, SARNET-2 spray benchmark

    Malet, J., E-mail: [Institut de Radioprotection et de Sûreté Nucléaire, Saclay (France); Mimouni, S., E-mail: [Electricité de France, EDF MF2E, Chatou (France); Manzini, G., E-mail: [RSE, Milano (Italy); Xiao, J., E-mail: [IKET, KIT, Karlsruhe (Germany); Vyskocil, L., E-mail: [UJV Rez (Czech Republic); Siccama, N.B., E-mail: [NRG, Safety and Power (Netherlands); Huhtanen, R., E-mail: [VTT, PO Box 1000, FI-02044 VTT (Finland)


    Highlights: • This paper presents a benchmark performed in the frame of the SARNET-2 EU project. • It concerns momentum transfer between a PWR spray and the surrounding gas. • The entrained gas velocities can vary up to 100% from one code to another. • Simplified boundary conditions for sprays are generally used by the code users. • It is shown how these simplified conditions impact the gas entrainment. - Abstract: This paper presents a benchmark performed in the frame of the SARNET-2 EU project, dealing with momentum transfer between a real-scale PWR spray and the surrounding gas. It presents a description of the IRSN tests on the CALIST facility, the participating codes (8 contributions), code-experiment and code-to-code comparisons. It is found that droplet velocities are almost well calculated one meter below the spray nozzle, even if the spread of the spray is not recovered and the values of the entrained gas velocity vary up to 100% from one code to another. Concerning sensitivity analysis, several ‘simplifications’ have been made by the contributors, especially based on the boundary conditions applied at the location where droplets are injected. It is shown here that such simplifications influence droplet and entrained gas characteristics. The next step will be to translate these conclusions in terms of variables representative of interesting parameters for nuclear safety.

  9. Surface Oxidation Phenomena of Ni-Based Alloy 600 in PWR Primary Water Conditions

    Lim, Yun Soo; Hwang, Seong Sik; Kim, Sung Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    There is, nevertheless, growing evidence in support for the internal oxidation model by Scot, in which grain boundary oxidation is responsible for embrittlement and cracking. Grain boundaries can act as an enhanced diffusion path for oxidation, and grain boundary oxidation can be regarded as a precursor for crack initiation. Oxidation of the grain boundary in almost all nickel-based alloys exposed to primary water is known to be detrimental for grin boundary cohesion. Panter et al. showed that the crack initiation time is strongly reduced when the specimens are pre-exposed in a simulated PWR environment in the absence of applied stress. The changes of the grain boundary structure and chemistry owing to oxygen penetration can increase the sensitivity to PWSCC under a load since grain boundary oxidization significantly weakens the grain boundary strength. Most of the important experimental results obtained are believed to correlate with the oxidation penetration into the material. A spinel structure was detected by XRD in the oxide layers. Several different types of oxide scales were found by SEM examination on the corroded surface of Alloy 600 after an immersion test in the primary water environments. Surface grain boundaries were oxidized by oxygen penetration into the matrix through grain boundaries. Grain boundary oxidization is thought to be the main reason for intergranular cracking in this alloy in a primary water environment of a PWR.

  10. On-line Autonomous Learning Based on Leamerg Expectation



    On-line autonomous learning of College English is one of the important reforms in colleges recently.This paper aims to explore the changes of teachers'role in the new on-line setting.The article first reviews the theoretical study of learner autonomy,then makes a practical investigation into the attitude and expectation learners have on teachers through a self-designed questionnaire,and explores that teachers should make an adjustment to their role orientation and changes their roles into motivators,evaluators and resources supphers in the new on-line setting.

  11. On-line Classical Guitar Course: Blogs for Music Education

    José Luis Navarro; Gilles Lavigne; G. Guadalupe Martínez Salgado


    This article introduces an on-line course constructed by means of a blog. The tool was the main goal of a research project titled “Develop, Implementation and evaluation of a Hybrid Course Face to face-On Line for Teaching the Beginning to Play the Classical Guitar”. This work was a three steep project in which it was implemented, applied and evaluated. The on-line course was intended to prepare the students to learn the basic principles to start in classical music with the guitar. The result...

  12. Understanding on-line community: the affordances of virtual space

    Karen Ruhleder


    Full Text Available Increasing numbers of on-line venues for learning are emerging as virtual communities become more accessible and commonplace. This paper looks at one particular virtual community, an on-line degree programme at the University of Illinois, Urbana-Champaign, which offers an M.S. in Library and Information Science (called LEEP. It draws on a framework presented by Mynatt, et al. (1998, which provides a lens for talking about on-line community as a set of affordances. This framework is applied to illustrate the interactions, artefacts, and expectations that shape this community.

  13. Polymorphism of GnRHR Gene and Its Relationship with Litter Size Trait of Gansu Meat Sheep New Breed Population%甘肃肉羊新品种选育群GnRHR基因多态性及其与产羔性状关联分析



    根据GenBank发布的绵羊促性腺激素释放激素受体(gonadotropin releasing hormone receptor,GnRHR)基因序列设计2对引物,采用聚合酶链式反应—单链构象多态性(PCR-SSCP)技术分析GnRHR基因外显子2和3在甘肃肉羊新品种选育群中的单核苷酸多态性,并与产羔性状进行关联分析.结果表明,在甘肃肉羊新品种选育群GnRHR基因外显子2检测到GG、GH、HH基因型,基因型频率分别为0.784 (160)、0.206( 42)、0.010(2),序列测序结果发现在编码区第198位碱基发生突变G→C,导致G变成R(甘氨酸→精氨酸)变化;GnRHR基因外显子3检测到MM、MN基因型,基因型频率分别为0.882(180)、0.118(24),序列测序结果发现在第257位碱基发生突变A→G,导致Q变成R(谷氨酰胺→精氨酸).GnRHR基因外显子2突变对新品种群羊繁殖力高低影响极显著,突变纯合基因型(HH)绵羊平均产羔数比野生纯合型(GG)多0.981只(P= 0.006<0.01);外显子3突变对新品种群羊繁殖力高低也有显著影响,MN基因型羊平均产羔数比MM基因型多0.688只(P=0.029<0.05).由此可推测GnRHR基因可能是控制新品种群羊产羔性能的一个主效基因或是与之存在紧密连锁的一个遗传标记.%The GnRHR gene was studied as a candidate gene for the high prolificacy of sheep. Two pairs of primers were designed according to the sequence of exon 2 and exon 3 of GnRHR gene published in Genbank.Single nucleotide polymorphisms ( SNPs )of exon 2 and exon 3 of GnRHR gene were detected in Gansu meat sheep new breed population and Poll Dorset by polymerase chain reaction (PCR)-single strand conformation polymorphism(SSCP). The results showed that there were three genotypes (GG,GH and HH)detected in GnRHR gene exon 2 of the new breed population, frequencies of GG,GH and HH genotypes were 0.784 (160),0.206 (42),0.010 (2),respectively. The polymorphic fragment amplified was sequenced and there was one nucleotide mutation (G→C)at c

  14. Research on General Corrosion Property of 304L and 304NG Stainless Steels in Simulated PWR Primary Water

    PENG; De-quan; HU; Shi-lin; ZHANG; Ping-zhu; WANG; Hui


    <正>The general corrosion behaviors of 304L and 304NG grade stainless steels in simulated pressurized water reactor (PWR) primary loop were studied using still autoclave, respectively, the corrosion test lasted for 1 680 hours. The corrosion oxide films were analyzed macroscopically and microscopically. The results are shown in Figs. 1, 2.

  15. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in BWR and PWR conditions

    Haenninen, H.; Aho-Mantila, I.


    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, has been examined both in O2-enriched BWR-conditions (8 ppm O2) and in typical PWR-conditions.

  16. On-Line Trajectory Retargeting for Alternate Landing Sites Project

    National Aeronautics and Space Administration — Barron Associates, Inc. proposes to develop a novel on-line trajectory optimization approach for Reusable Launch Vehicles (RLVs) under failure scenarios, targeting...


    Cuixia MIAO; Yuzhong ZHANG


    In this paper, we consider the on-line scheduling of unit time jobs with rejection on m identical parallel machines. The objective is to minimize the total completion time of the accepted jobs plus the total penalty of the rejected jobs. We give an on-line algorithm for the problem with competitive ratio 1/2(2 + √3) ≈ 1.86602.

  18. From User Comments to On-line Conversations

    Wang, Chunyan; Ye, Mao; Huberman, Bernardo A.


    We present an analysis of user conversations in on-line social media and their evolution over time. We propose a dynamic model that accurately predicts the growth dynamics and structural properties of conversation threads. The model successfully reconciles the differing observations that have been reported in existing studies. By separating artificial factors from user behaviors, we show that there are actually underlying rules in common for on-line conversations in different social media web...

  19. Directions for Future Research in On-line Distance Education

    Alaa SADIK


    Although institutions have invested much in developing on-line environments or using already established commercial platforms, only few studies have been conducted to investigate the effectiveness of on-line courses based on empirical data (Jung and Rha, 2000). A review of the literature conducted in this study showed that most of online learning studies investigated the effectiveness of Web-based interaction or Internet conferencing on learning, not the entire learning environment. Even in t...

  20. An On-line Ferrograph for Monitoring Machine Wear

    L(U) Xiao-jun; JING Min-qing; XIE You-bai


    In order to improve an on-line ferrograph, this paper simulates a three dimensional magnetic field distribution of an electromagnet, builds a sinking motion model of a wear particle, and investigates the motion law of wear particles under two different conditions. Both numeric results and experimental results show that the on-line ferrograph is capable of monitoring machine wear conditions by measuring the concentration and size distribution of wear particles in lubricating oil.

  1. Development of an artificial neural network model for on-line thermal margin estimation of a nuclear reactor core

    Kim, Hyun Koon


    One of the key safety parameters related to thermal margin in a Pressurized Water Reactor (PWR) core, is Departure from Nucleate Boiling Ratio (DNBR), which is to be assessed and continuously monitored during operation via either an analog or a digital monitoring system. The digital monitoring system, in general, allows more thermal margin than the analog system through the on-line computation of DNBR using the measured parameters as inputs to a simplified, fast running computer code. The purpose of this thesis is to develop an advanced method for on-line DNBR estimation by introducing an artifactual neural network model for best-estimation of DNBR at the given reactor operating conditions. the neural network model, consisting of three layers with five operating parameters in the input layer, provides real-time prediction accuracy of DNBR by training the network against the detailed simulation results for various operating conditions. The overall training procedure is developed to learn the characteristics of DNBR behaviour in the reactor core. First, a set of random combination of input variables is generated by Latin Hypercube Sampling technique performed on a wide range of input parameters. Second, the target values of DNBR to be referenced for training are calculated using a detailed simulation code, COBRA-IV. Third, the optimized training input data are selected. Then, training is performed using an Error Back Propagation algorithm. After completion of training, the network is tested on the examining data set in order to investigate the generalization capability of the network responses for the steady state operating condition as well as for the transient situations where DNB is of a primary concern. The test results show that the values of DNBR predicted by the neural network are maintained at a high level of accuracy for the steady state condition, and are in good agreements with the transient situation, although slightly conservative as compared to those

  2. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.


    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  3. Effect of Weld Properties on the Crush Strength of the PWR Spacer Grid

    Kee-nam Song


    Full Text Available Mechanical properties in a weld zone are different from those in the base material because of different microstructures. A spacer grid in PWR fuel is a structural component with an interconnected and welded array of slotted grid straps. Previous research on the strength analyses of the spacer grid was performed using base material properties owing to a lack of mechanical properties in the weld zone. In this study, based on the mechanical properties in the weld zone of the spacer grid recently obtained by an instrumented indentation technique, the strength analyses considering the mechanical properties in the weld zone were performed, and the analysis results were compared with the previous research.

  4. Sensitivity analysis of a PWR fuel element using zircaloy and silicon carbide claddings

    Faria, Rochkhudson B. de; Cardoso, Fabiano; Salome, Jean A.D.; Pereira, Claubia; Fortini, Angela, E-mail:, E-mail: [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Escola de Engenharia. Departamento de Engenharia Nuclear


    The alloy composed of zirconium has been used effectively for over 50 years in claddings of nuclear fuel, especially for PWR type reactors. However, to increase fuel enrichment with the aim of raising the burning and maintaining the safety of nuclear plants is of great relevance the study of new materials that can replace safely and efficiently zircaloy cladding. Among several proposed material, silicon carbide (SiC) has a potential to replace zircaloy as fuel cladding material due to its high-temperature tolerance, chemical stability and low neutron affinity. In this paper, the goal is to expand the study with silicon carbide cladding, checking its behavior when submitted to an environment with boron, burnable poison rods, and temperature variations. Sensitivity calculation and the impact in multiplication factor to both claddings, zircaloy and silicon carbide, were performed during the burnup. The neutronic analysis was made using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation) code. (author)

  5. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Hermann, O.W.


    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  6. Steady characteristic investigation on passive residual heat removal system of Chinese advanced PWR


    Thermal-hydraulic characteristic investigation on passive residual heat removal system(PRHRS)of Chinese advanced PWR was conducted to provide input data for PRHRS design and to demonstrate the feasibility of unique design features.A total of 237 sets of test data at steady state have been obtained and the main influence factors on the two-phase natural circulation flow rate and residual heat removal capability were identified.On the basis of theory analysis,a correlation of two-phase natural circulation was obtained,and relative errors of 95% test data were less than±16%.There is a considerable effect of the system status parameters on the threshold of height between heat source and heat sink,and its correlation of two-phase natural circulation system has been obtained.The steady characteristic research shows that PRHRS has the capability of removing the core decay power through natural circulation.

  7. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter


    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  8. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)


    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  9. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)


    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  10. Computational analysis for prediction of pressure of PWR presurizer undertransient conditions


    A computer model has been developed for prediction of the pressure in thepressurizer undertransient conditions. In the model three separate thermodynamic regions which arenot required to be inthermal equilibrium have been considered. The mathematical model derived from the general conservation equations includesall of theimportant thermal-hydraulics phenomena occurring in the pressurizer,i.e., stratificationof the hot water andincoming cold water, bulk flashing and condensation, wall condensation, andinterfacial heat and masstransfer, etc. The bubble rising and rain-out models are developed to describe bulkflashing andcondensation, respectively. To obtain the wall condensation rate, a one-dimensionalheat conductionequation is solved by the pivoting method. The presented model will predict thepressure-time behaviorof a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreementwithavailable data on the CHASHMA nuclear power plant's pressurizer performance.


    阿谢德; 徐济鋆


    A computer program PRETTA “Pressurizer Transient Thermodynamics Analysis” was developed for the prediction of pressurizer under transient conditions. It is based on the solution of the conservation laws of heat and mass applied to the three separate and non equilibrium thermodynamic regions. In the program all of the important thermal-hydraulics phenomena occurring in the pressurizer: stratification of the hot water and incoming cold water, bulk flashing and condensation, wall condensation, and interfacial heat and mass transfer have been considered. The bubble rising and rain-out models are developed to describe bulk flashing and condensation, respectively. To obtain the wall condensation rate, a one-dimensional heat conduction equation is solved by the pivoting method. The presented computer program will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The results obtained from the proposed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant's pressurizer performance.

  12. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas


    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  13. Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

    Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10{sup 4} to 3.85 x 10{sup 5}. (Author) 5 refs., 18 figs., 5 tabs.

  14. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.


    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  15. Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code

    Croff, A. G.; Bjerke, M. A.; Morrison, G. W.; Petrie, L. M.


    Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given.

  16. Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

    Fujishiro, Toshio; Yanagisawa, Kazuaki; Ishijima, Kiyomi; Shiba, Koreyuki


    Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

  17. Quantitative uncertainty and sensitivity analysis of a PWR control rod ejection accident

    Pasichnyk, I.; Perin, Y.; Velkov, K. [Gesellschaft flier Anlagen- und Reaktorsicherheit - GRS mbH, Boltzmannstasse 14, 85748 Garching bei Muenchen (Germany)


    The paper describes the results of the quantitative Uncertainty and Sensitivity (U/S) Analysis of a Rod Ejection Accident (REA) which is simulated by the coupled system code ATHLET-QUABOX/CUBBOX applying the GRS tool for U/S analysis SUSA/XSUSA. For the present study, a UOX/MOX mixed core loading based on a generic PWR is modeled. A control rod ejection is calculated for two reactor states: Hot Zero Power (HZP) and 30% of nominal power. The worst cases for the rod ejection are determined by steady-state neutronic simulations taking into account the maximum reactivity insertion in the system and the power peaking factor. For the U/S analysis 378 uncertain parameters are identified and quantified (thermal-hydraulic initial and boundary conditions, input parameters and variations of the two-group cross sections). Results for uncertainty and sensitivity analysis are presented for safety important global and local parameters. (authors)

  18. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A


    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  19. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail:, E-mail:, E-mail: [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)


    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  20. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay


    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  1. Fatigue Crack Growth Rate of Type 347 Stainless Steel at the PWR Environment

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Materials used in nuclear power plants are low alloy steel, stainless steel, and superalloy steel. Understanding the characteristics of these materials is important in the development of nuclear power plant related technology. Nb-stabilized Type 347 stainless steel is used for the coolant pressurizer surge line of Korea Standard Nuclear Power Plant (KSNPP). Surge line of PWR nuclear reactor are damaged by thermal fatigue due to thermal gradient during heat-up and cool-down, mechanical fatigue due to mechanical stress, and corrosion fatigue due to nuclear reactor water environment. Fatigue is an important factor which limits the life of structure. Fatigue crack growth rate curves in nuclear reactor environment are needed to evaluate the integrity of nuclear reactor structure but that result is not sufficient. In this study, fatigue crack growth rates at nuclear reactor environment are produced to evaluate integrity of nuclear power plant section 5

  2. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    Kelly, G N; Charles, D; Hemming, C R


    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

  3. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)


    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  4. Sport Management Taught On-Line: A Discussion

    William F. Stier Jr


    Full Text Available An introduction to the world of on-line courses (distance education/learning is presented. In addition, the world of on-line learning, as it pertains to sport management, is examined within the framework of (a pedagogy, (b finances,(c assessment, and (d choosing to transition from the traditional classroom to on-line learning. Pertinent points relative to each of the four categories are presented from the literature. In an effort to stimulate thought and discussion to the subject of on-line learning for sport management programs/courses the authors provide their reactions to the literature points by presenting their comments/reactions from a sport management perspective. Sport management professors and administrators are encouraged to critically examine the feasibility of such on-line courses (distance education/learning within their own curricula while maintaining an appropriate framework revolving around sound theoretical instructional strategies, methods as well as appropriate use of instructional tools, including but not limited to, computersand the WWW.

  5. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    DeHart, M.D.


    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  6. Developing and analyzing long-term fuel management strategies for an advanced Small Modular PWR

    Hedayat, Afshin, E-mail:


    Highlights: • Comprehensive introduction and supplementary concepts as a review paper. • Developing an integrated long-term fuel management strategy for a SMR. • High reliable 3-D core modeling over fuel pins against the traditional LRM. • Verifying the expert rules of large PWRs for an advanced small PWR. • Investigating large numbers of safety parameters coherently. - Abstract: In this paper, long-term fuel management (FM) strategies are introduced and analyzed for a new advanced Pressurized Light Water Reactor (PWR) type of Small Modular Reactors (SMRs). The FM strategies are developed to be safe and practical for implementation as much as possible. Safety performances, economy of fuel, and Quality Assurance (QA) of periodic equilibrium conditions are chosen as the main goals. Flattening power density distribution over fuel pins is the major method to ensure safety performance; also maximum energy output or permissible discharging burn up indicates economy of fuel fabrication costs. Burn up effects from BOC to EOC have been traced, studied, and highly visualized in both of transport lattice cell calculations and diffusion core calculations. Long-term characteristics are searched to gain periodical equilibrium characteristics. They are fissile changes, neutron spectrum, refueling pattern, fuel cycle length, core excess reactivity, average, and maximum burn up of discharged fuels, radial Power Peaking Factors (PPF), total PPF, radial and axial power distributions, batch effects, and enrichment effects for fine regulations. Traditional linear reactivity model have been successfully simulated and adapted via fine core and burn up calculations. Effects of high burnable neutron poison and soluble boron are analyzed. Different numbers of batches via different refueling patterns have been studied and visualized. Expert rules for large type PWRs have been influenced and well tested throughout accurate equilibrium core calculations.

  7. Evaluation of Physical Characteristics of PWR Cores with Accident Tolerant Fuels

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)


    The accident tolerant fuels (ATF) considered in this work includes metallic microcell UO{sub 2} pellets and outer Cr-based alloy coating on cladding, which is being developed in KAERI (Korea Atomic Energy Research Institute). Chromium metals have been used in many fields because of its hardness and corrosion-resistance. The use of the chromium metal in nuclear fuel rod can enhance the conductivity of pellets and corrosion-resistance of cladding. The objective of this work is to study the neutronic performances and characteristics of the commercial PWR core loaded the ATF-bearing assemblies. In this work, we studied the PWR cores which are loaded with ATF assemblies to improve the safety of reactor core. The ATF rod consists of the metallic microcell UO2 pellet which includes chromium of 3.34 wt% and the outer 0.05mm thick coating of Cr-based alloy with atomic number ratio of 85:15. We performed the cycle-by-cycle reload core analysis from the cycle 8 at which the ATF fuel assemblies start to be loaded into the core. The target nuclear power plant is the Hanbit-3 nuclear power plant. From the analysis, it was found that 1) the uranium enrichment is required to be increased up to 5.20/4.70 wt% in order to satisfy a required cycle length of 480 EFPDs, 2) the cycle length for the core using ATF fuel assemblies with the same uranium enrichments as those in the reference UO{sub 2} fueled core is decreased from 480 EFPDs to 430 EFPDs.

  8. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    Morlent, O.; Reuchet, J. [CEA Fontenay-aux-Roses, Inst. de Protection et de Surete Nucleaire, 92 (France)


    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  9. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)


    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  10. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Kang, Shin K., E-mail:; Hassan, Yassin A.


    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  11. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W


    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  12. Effect of aging on the PWR Chemical and Volume Control System

    Grove, E.J.; Travis, R.J.; Aggarwal, S.K. [Brookhaven National Lab., Upton, NY (United States)


    The PWR Chemical and Volume Control System (CVCS) is designed to provide both safety and non-safety related functions. During normal plant operation it is used to control reactor coolant chemistry, and letdown and charging flow. In many plants, the charging pumps also provide high pressure injection, emergency boration, and RCP seal injection in emergency situations. This study examines the design, materials, maintenance, operation and actual degradation experiences of the system and main sub-components to assess the potential for age degradation. A detailed review of the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Report (LER) databases for the 1988--1991 time period, together with a review of industry and NRC experience and research, indicate that age-related degradations and failures have occurred. These failures had significant effects on plant operation, including reactivity excursions, and pressurizer level transients. The majority of these component failures resulted in leakage of reactor coolant outside the containment. A representative plant of each PWR design (W, CE, and B and W) was visited to obtain specific information on system inspection, surveillance, monitoring, and inspection practices. The results of these visits indicate that adequate system maintenance and inspection is being performed. In some instances, the frequencies of inspection were increase in response to repeated failure events. A parametric study was performed to assess the effect of system aging on Core Damage Frequency (CDF). This study showed that as motor-operated valve (MOV) operating failures increased, the contribution of the High Pressure Injection to CDF also increased.

  13. On-line Payment System Survey – eCash

    Marius Popa


    Full Text Available The paper presents the main aspects regarding an on-line payment system. Some characteristics of such system are presented and an existing system is analyzed. On its fundamental sense, the electronic commerce is a concept that represents the purchase and sale process or exchange of products, services, information, using o computer network, inclusively the Internet. In the most part of the cases, the electronic commerce imply on-line payments that lead to creation of some kinds of electronic money and some specific payment systems. There are described the some electronic payment mechanisms and the architecture and the functions of the on-line payment system E-Cash are depicted.

  14. A Hybrid On-line Verification Method of Relay Setting

    Gao, Wangyuan; Chen, Qing; Si, Ji; Huang, Xin


    Along with the rapid development of the power industry, grid structure gets more sophisticated. The validity and rationality of protective relaying are vital to the security of power systems. To increase the security of power systems, it is essential to verify the setting values of relays online. Traditional verification methods mainly include the comparison of protection range and the comparison of calculated setting value. To realize on-line verification, the verifying speed is the key. The verifying result of comparing protection range is accurate, but the computation burden is heavy, and the verifying speed is slow. Comparing calculated setting value is much faster, but the verifying result is conservative and inaccurate. Taking the overcurrent protection as example, this paper analyses the advantages and disadvantages of the two traditional methods above, and proposes a hybrid method of on-line verification which synthesizes the advantages of the two traditional methods. This hybrid method can meet the requirements of accurate on-line verification.

  15. On-line laser spectroscopy with thermal atomic beams

    Thibault, C; De Saint-Simon, M; Duong, H T; Guimbal, P; Huber, G; Jacquinot, P; Juncar, P; Klapisch, Robert; Liberman, S; Pesnelle, A; Pillet, P; Pinard, J; Serre, J M; Touchard, F; Vialle, J L


    On-line high resolution laser spectroscopy experiments have been performed in which the light from a CW tunable dye laser interacts at right angles with a thermal atomic beam. /sup 76-98/Rb, /sup 118-145 /Cs and /sup 208-213/Fr have been studied using the ionic beam delivered by the ISOLDE on-line mass separator at CERN while /sup 30-31/Na and /sup 38-47/K have been studied by setting the apparatus directly on-line with the PS 20 GeV proton beam. The principle of the method is briefly explained and some results concerning nuclear structure are given. The hyperfine structure, spins and isotope shifts of the alkali isotopes and isomers are measured. (8 refs).

  16. Strengthening weak ties through on-line gaming

    Sudzina, Frantisek; Razmerita, Liana Virginia; Kirchner, Kathrin

    On-line gaming became widespread in the last couple of years. The aim of the research presented in the paper is to figure out to what extent does game playing helps to strengthen weak ties and what additional factors influence this process. The approach is rather exploratory – some factors...... are grounded in theory, some are new. These factors are age, gender, place of origin, number of their Facebook connections (friends in Facebook terminology), the amount of time they are on Facebook, the amount of time they keep the Facebook site open, the amount of time they play on-line games, and the reasons...... for starting to play on-line games. Regarding the latter, we chose to focus only on escapist reasons....

  17. On-Line Voltage Stability Assessment based on PMU Measurements

    Garcia-Valle, Rodrigo; P. Da Silva, Luiz C.; Nielsen, Arne Hejde


    through statistic analysis. During the off-line analysis, a memory of high-risk situations following a pre-defined voltage stability criterion is obtained. Thereafter, basic statistics analyses are applied resulting in the definition of voltage regions. During on-line operation, voltage magnitudes......This paper presents a method for on-line monitoring of risk voltage collapse based on synchronised phasor measurement. As there is no room for intensive computation and analysis in real-time, the method is based on the combination of off-line computation and on-line monitoring, which are correlated...... of critical buses obtained by phasor measurements are monitored in relation to the risk regions. Comprehensive studies demonstrate that the proposed method could assist operators to avoid voltage collapse events, by taking preventive or emergency actions....

  18. On-line tribochemical strengthening of gear surface


    It has been found that under favorable friction conditions some antiwear elements inlubricating additives can permeate into subsurface of metal which can strengthens the friction sur-faces and improves anti-wear capacity of frication pairs. It is in many ways similar to chemical heattreatment. A new concept, technology of on-line strengthening, was logically put forward. Based oncurrent gear surface treatment technology, the on-line strengthen of gear surface is proposed. Itsdesign method is established. Based on it, the on-line strengthen of gear is achieved on CL-100gear test machine. A new method is put forward for strengthen treatment of gear surface. Andthree kinds of surface film were suggested.

  19. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki


    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  20. Why do People Stop Playing On-Line Games?

    Sudzina, Frantisek; Razmerita, Liana


    The recent initial public offering of shares of Zynga, probably the most important on-line game provider, drew interest of potential investors but also of general public to their business model. What the most interested people learned so far is that if Zynga had not changed their accounting...... practice, they would be in red numbers for several months already. This is most likely caused by people stopping to play their games. This paper provides an estimate of what proportion of people, who played on-line games, already stopped playing them. Additionally, it analyzed the reasons why people...

  1. On-Line Generation of 3D-Waves

    Frigaard, Peter


    The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg.......The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg....

  2. The User-friendly On-Line Diffusion Chamber

    Aviles Acosta, Jaime


    The On-Line Diffusion Chamber is a stand-alone apparatus built to carry out short-live radiotracer diffusion studies. The availability of the on-demand production of isotopes in the ISOLDE facility, and the design of the apparatus to streamline the implantation process, annealing treatment, ion gun ablation with a tape transport system, and radiation intensity measurement with a Ge gamma detector all in the same apparatus, gives the On-Line Diffusion Chamber a unique ability to studies with short-lived radioisotopes or isomer states that are not possible in any other facility in the world.

  3. On-line measurement of heat of combustion

    Chaturvedi, S. K.; Chegini, H.


    An experimental method for an on-line measurement of heat of combustion of a gaseous hydrocarbon fuel mixture of unknown composition is developed. It involves combustion of a test gas with a known quantity of air to achieve a predetermined oxygen concentration level in the combustion products. This is accomplished by a feedback controller which maintains the gas volumetric flow rate at a level consistent with the desired oxygen concentration in the products. The heat of combustion is determined from a known correlation with the gas volumetric flow rate. An on-line microcomputer accesses the gas volumetric flow data, and displays the heat of combustion values at desired time intervals.

  4. 5th Computer Science On-line Conference

    Senkerik, Roman; Oplatkova, Zuzana; Silhavy, Petr; Prokopova, Zdenka


    This volume is based on the research papers presented in the 5th Computer Science On-line Conference. The volume Artificial Intelligence Perspectives in Intelligent Systems presents modern trends and methods to real-world problems, and in particular, exploratory research that describes novel approaches in the field of artificial intelligence. New algorithms in a variety of fields are also presented. The Computer Science On-line Conference (CSOC 2016) is intended to provide an international forum for discussions on the latest research results in all areas related to Computer Science. The addressed topics are the theoretical aspects and applications of Computer Science, Artificial Intelligences, Cybernetics, Automation Control Theory and Software Engineering.

  5. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)


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

  6. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.


    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  7. Development of a model of a NSSS of the PWR reactor with thermo-hydraulic code GOTHIC; Desarrollo de un modelo del NSSS de un reactor PWR con el codigo termo-hidraulico GOTHIC

    Gomez Garcia-Torano, I.; Jimenez, G.


    The Thermo-hydraulic code GOTHIC is often used in the nuclear industry for licensing transient analysis inside containment of generation II (PWR, BWR) plants as Gen III and III + (AP1000, ESBWR, APWR). After entering the mass and energy released to the containment, previously calculated by other codes (basis, TRACE), GOTHIC allows to calculate in detail the evolution of basic parameters in the containment.

  8. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Fernandez, J. L.; Lopez, J.


    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  9. Qualitative analysis of the maintenance politics of the systems of a typical PWR by artificial neural networks; Analise qualitativa da politica de manutencoes dos sistemas de um PWR tipico por redes neurais artificiais

    Lourenco, Victor Hugo Moreno


    Proceedings and techniques in order to maximize the reliability and the availability of industrial plants have been used along the last decades by specialists and professionals of maintenance. However, the modem industrial systems' sizing, and the increasing complexity and interdependence among its components have become this activity's planning a more and more difficult task. Considering this scenario, the objective of the present work is to provide a computational tool which is able to help about the taking decision's task, and about planning policies of maintenance practiced in thermonuclear plants. The tool developed is based on the artificial neural networks (ANN) for the recognition of standards and establishment of correlations among events occurred in the components of pressurized water reactor (PWR) typical systems. The ANN work as miners of database of failure events, and are able to identify connections and to establish imperceptible inferences even for the most experienced specialists in maintenance of nuclear systems. The results were attained from realistic data and are confronted against the maintenance's classic policies which are practiced nowadays on PWR thermonuclear plants. These results show the solidity of the technique in valuing and predicting failures in a real power plant, and is able to be used as a tool for supporting decisions about planning maintenance policies on a typical PWR. (author)

  10. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Cheverton, R.D.


    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  11. Application of RELAP5/MOD1 for calculation of safety and relief valve discharge piping hydrodynamic loads. Final report. [PWR


    A series of operability tests of spring-loaded safety valves was performed at Combustion Engineering in Windsor, CT as part of the PWR Safety and Relief Valve Test Program conducted by EPRI on behalf of PWR Utilities in response to the recommendations of NUREG-0578 and the requirements of the NRC. Experimental data from five of the safety valve tests are compared with RELAP5/MOD1 calculations to evaluate the capability of the code to determine the fluid-induced transient loads on downstream piping. Comparisons between data and calculations are given for transients with discharge of steam, water, and water loop seal followed by steam. RELAP5/MOD1 provides useful engineering estimates of the fluid-induced piping loads for all cases.

  12. Calculation of sample problems related to two-phase flow blowdown transients in pressure relief piping of a PWR pressurizer

    Shin, Y.W.; Wiedermann, A.H.


    A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients.

  13. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Oettingen Mikołaj


    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  14. PFM Analysis for Pre-Existing Cracks on Alloy 182 Weld in PWR Primary Water Environment using Monte Carlo Simulation

    Park, Jae Phil; Bahn, Chi Bum [Pusan National University, Busan (Korea, Republic of)


    Probabilistic Fracture Mechanics (PFM) analysis was generally used to consider the scatter and uncertainty of parameters in complex phenomenon. Weld defects could be present in weld regions of Pressurized Water Reactors (PWRs), which cannot be considered by the typical fracture mechanics analysis. It is necessary to evaluate the effects of the pre-existing cracks in welds for the integrity of the welds. In this paper, PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out using a Monte Carlo simulation. PFM analysis for pre-existing cracks on Alloy 182 weld in PWR primary water environment was carried out. It was shown that inspection decreases the gradient of the failure probability. And failure probability caused by the pre-existing cracks was stabilized after 15 years of operation time in this input condition.

  15. On-Line Learning and the Implications for School Design

    Stack, Greg


    "Disrupting Class," published in 2008, is the story of how disruptive innovation, innovation that changes the business model organizations, will fundamentally change the American school system. The book's most startling prediction is that half of all high school classes will be on-line by 2019. In considering these predictions, the author began to…

  16. On-line multidimensional separation systems for peptide analysis

    Stroink, T.


    Today, there is an increasing interest in selective and sensitive analysis of proteins and peptides with a relatively high speed. The first chapter of this thesis describes several strategies for the on-line multidimensional analysis of peptides and proteins in biological samples. This overview of t

  17. On-line probabilistic classification with particle filters

    Højen-Sørensen, Pedro; de Freitas, N.; Fog, Torben L.


    We apply particle filters to the problem of on-line classification with possibly overlapping classes. This allows us to compute the probabilities of class membership as the classes evolve. Although we adopt neural network classifiers, the work can be extended to any other parametric classification...

  18. Personal Assistant for onLine Services: Addressing human factors

    Lindenberg, J.; Nagata, S.F.; Neerincx, M.A.


    The Personal Assistant for onLine Services (PALS) project aims at substantially improving the user experience of mobile internet services. It focuses on a generic solution: a personal assistant, which attunes the interaction to the momentary user needs and use context (e.g. adjusting the

  19. Efficiently Building On-line Tools for Distributed Heterogeneous Environments

    Günther Rackl


    Full Text Available Software development is getting more and more complex, especially within distributed middleware-based environments. A major drawback during the overall software development process is the lack of on-line tools, i.e. tools applied as soon as there is a running prototype of an application. The MIMO MIddleware MOnitor provides a solution to this problem by implementing a framework for an efficient development of on-line tools. This paper presents a methodology for developing on-line tools with MIMO. As an example scenario, we choose a distributed medical image reconstruction application, which represents a test case with high performance requirements. Our distributed, CORBA-based application is instrumented for being observed with MIMO and related tools. Additionally, load balancing mechanisms are integrated for further performance improvements. As a result, we obtain an integrated tool environment for observing and steering the image reconstruction application. By using our rapid tool development process, the integration of on-line tools shows to be very convenient and enables an efficient tool deployment.

  20. Developing an On-Line Interactive Health Psychology Module

    Upton, Dominic; Cooper, Carol


    On-line teaching material in health psychology was developed which ensured a range of students could access appropriate material for their course and level of study. This material has been developed around the concept of smaller "content chunks" which can be combined into whole units of learning (topics), and ultimately, a module. On the basis of…

  1. Why do People Stop playing On-Line Games?

    Sudzina, Frantisek; Razmerita, Liana


    The recent initial public offering of shares of Zynga, probably the most important on-line game provider, drew interest of potential investors but also of general public to their business model. What the most interested people learned so far is that if Zynga had not changed their accounting...

  2. On-line fuzzy logic control of tube bending

    Lieh, Junghsen; Li, Wei Jie


    This paper describes the simulation and on-line fuzzy logic control of tube bending. By combining elasticity and plasticity theories, a conventional model was developed. The results from simulation were compared with those obtained from testing. The experimental data reveal that there exists certain level of uncertainty and nonlinearity in tube bending, and its variation could be significant. To overcome this, a on-line fuzzy logic controller with self-tuning capabilities was designed. The advantages of this on-line system are (1) its computational requirement is simple in comparison with more algorithmic-based controllers, and (2) the system does not need prior knowledge of material characteristics. The device includes an AC motor, a servo controller, a forming mechanism, a 3D optical sensor, and a microprocessor. This automated bending machine adopts primary and secondary errors between the actual response and desired output to conduct on-line rule reasoning. Results from testing show that the spring back angle can be effectively compensated by the self- tuning fuzzy system in a real-time fashion.

  3. The dynamics of on-line principal component analysis

    Biehl, M.; Schlösser, E.


    The learning dynamics of an on-line algorithm for principal component analysis is described exactly in the thermodynamic limit by means of coupled ordinary differential equations for a set of order parameters. It is demonstrated that learning is delayed significantly because existing symmetries amon

  4. On-Line Learning and the Implications for School Design

    Stack, Greg


    "Disrupting Class," published in 2008, is the story of how disruptive innovation, innovation that changes the business model organizations, will fundamentally change the American school system. The book's most startling prediction is that half of all high school classes will be on-line by 2019. In considering these predictions, the author began to…

  5. Specialization processes in on-line unsupervised learning

    Biehl, M.; Freking, A.; Reents, G.; Schlösser, E.


    From the recent analysis of supervised learning by on-line gradient descent in multilayered neural networks it is known that the necessary process of student specialization can be delayed significantly. We demonstrate that this phenomenon also occurs in various models of unsupervised learning. A sol

  6. On-line sample treatment - Capillary gas chromatography

    Goosens, EC; de Jong, D; de Jong, GJ; Brinkman, UAT


    Sample pretreatment is often the bottleneck of a trace level analytical procedure. In order to increase performance, increasing attention is therefore being devoted to combining sample pretreatment on-line with the separation technique that has to be used. In the present review, a variety of procedu

  7. A New On-Line Resource for Psycholinguistic Studies

    Szekely, Anna; Jacobsen, Thomas; D'Amico, Simona; Devescovi, Antonella; Andonoa, Elena; Herron, Daniel; Lu, Ching Ching; Pechmann, Thomas; Pleh, Csaba; Wicha, Nicole; Federmeier, Kara; Gerdjikova, Irina; Gutierrez, Gabriel; Hung, Daisy, Hsu, Jeanne; Iyer, Gowri; Kohnert, Kathryn; Mehotcheva, Teodora; Orozco-Figueroa, Araceli; Tzeng, Angela; Tzeng, Ovid; Arevalo, Analia; Vargha, Andras; Butler, Andrew C.; Buffington, Robert; Bates, Elizabeth


    Picture naming is a widely used technique in psycholinguistic studies. Here, we describe new on-line resources that our project has compiled and made available to researchers on the world wide web at The website provides access to a wide range of picture stimuli and related norms in seven languages. Picture…


    Iwayama, N.; Ishigaki, K.


    We propose a new approach to context processing in on-line handwritten character recognition (OLCR). Based on the observation that writers often repeat the strings that they input, we take the approach of adaptive context processing. (ACP). In ACP, the strings input by a writer are automatically

  9. The Lesson Observation On-Line (Evidence Portfolio) Platform

    Cooper, David G.


    At a time when teacher training is being moved to school-based programmes it is important to engage in a research-informed dialogue about creating more distinctive, and cost-effective 21st century models of teacher training. Three years ago I began feasibility field testing the Lesson Observation On-line (Evidence Portfolio) Platform [LOOP]…

  10. On-line EM algorithm for the normalized gaussian network.

    Sato, M; Ishii, S


    A normalized gaussian network (NGnet) (Moody & Darken, 1989) is a network of local linear regression units. The model softly partitions the input space by normalized gaussian functions, and each local unit linearly approximates the output within the partition. In this article, we propose a new on-line EMalgorithm for the NGnet, which is derived from the batch EMalgorithm (Xu, Jordan, &Hinton 1995), by introducing a discount factor. We show that the on-line EM algorithm is equivalent to the batch EM algorithm if a specific scheduling of the discount factor is employed. In addition, we show that the on-line EM algorithm can be considered as a stochastic approximation method to find the maximum likelihood estimator. A new regularization method is proposed in order to deal with a singular input distribution. In order to manage dynamic environments, where the input-output distribution of data changes over time, unit manipulation mechanisms such as unit production, unit deletion, and unit division are also introduced based on probabilistic interpretation. Experimental results show that our approach is suitable for function approximation problems in dynamic environments. We also apply our on-line EM algorithm to robot dynamics problems and compare our algorithm with the mixtures-of-experts family.

  11. On-Line Synthesis and Analysis by Mass Spectrometry

    Bain, Ryan M.; Pulliam, Christopher J.; Raab, Shannon A.; Cooks, R. Graham


    In this laboratory experiment, students learn how to use ESI to accelerate chemical synthesis and to couple it with on-line mass spectrometry for structural analysis. The Hantzsch synthesis of symmetric 1,4-dihydropyridines is a classic example of a one-pot reaction in which multiple intermediates can serve to indicate the progress of the reaction…

  12. Investigating on-line pornography at the University of Johannesburg

    P. Laughton


    Full Text Available The on-line user of today has access to a vast collection of information resources. In addition, the developments in Internet and Web technologies have made it even easier for surfers to anonymously get access to on-line pornography. The purpose of this research was to investigate the extent to which access to on-line pornography at the University of Johannesburg can be managed. For the empirical part of this research 1037 questionnaires were proportionally distributed to and completed by students on all five campuses of the university. The questionnaire consisted of four sections: biographical information; university computer facility usage; university acceptable use policy; and personal experience with university computer facilities. The gender distribution for the sample was almost even, with a total of 49,4% male participants and 50,6% female, with the largest grouping of respondents (61,6% aged between 19 years and 21 years. Of the respondents, 36,7% indicated that exposure to unsolicited pornography did not bother them. When asked to what extent students should have access to pornography, 60,5% stated 'None' while 32,6% believed that 'Restricted' access should be granted for research purposes and 6,9% believed that students should be granted 'Total' access to pornography. Results from the research will be used to manage access to on-line resources at the University of Johannesburg better.

  13. Investigating on-line pornography at the University of Johannesburg

    P. Laughton


    Full Text Available The on-line user of today has access to a vast collection of information resources. In addition, the developments in Internet and Web technologies have made it even easier for surfers to anonymously get access to on-line pornography. The purpose of this research was to investigate the extent to which access to on-line pornography at the University of Johannesburg can be managed. For the empirical part of this research 1037 questionnaires were proportionally distributed to and completed by students on all five campuses of the university. The questionnaire consisted of four sections: biographical information; university computer facility usage; university acceptable use policy; and personal experience with university computer facilities. The gender distribution for the sample was almost even, with total of 49,4% male participants and 50,6% female, with the largest grouping of respondents(61,6% aged between 19 years and 21 years. Of the respondents, 36,7% indicated that exposure to unsolicited pornography did not bother them. When asked to what extent students should have access to pornography, 60,5% stated 'None' while 32,6% believed that 'Restricted' access should be granted for research purposes and 6,9% believed that students should be granted 'Total' access to pornography. Results from the research will be used to manage access to on-line resources at the University of Johannesburg better.

  14. On-Line NDE for Advanced Reactor Designs

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.


    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  15. On-Line Synthesis and Analysis by Mass Spectrometry

    Bain, Ryan M.; Pulliam, Christopher J.; Raab, Shannon A.; Cooks, R. Graham


    In this laboratory experiment, students learn how to use ESI to accelerate chemical synthesis and to couple it with on-line mass spectrometry for structural analysis. The Hantzsch synthesis of symmetric 1,4-dihydropyridines is a classic example of a one-pot reaction in which multiple intermediates can serve to indicate the progress of the reaction…

  16. Why do People Stop Playing On-Line Games?

    Sudzina, Frantisek; Razmerita, Liana


    The recent initial public offering of shares of Zynga, probably the most important on-line game provider, drew interest of potential investors but also of general public to their business model. What the most interested people learned so far is that if Zynga had not changed their accounting...

  17. On-line survivable routing in WDM networks

    Beshir, A.A.; Kuipers, F.A.; Van Mieghem, P.F.A.; Orda, A.


    In WDM networks, survivable routing and wavelength assignment (SRWA) involves assigning link-disjoint primary and backup lightpaths. In the on-line SRWA problem, a sequence of requests arrive and each request is either accepted or rejected based only on the input sequence seen so far. For special

  18. On-line Survivable Routing in WDM Networks

    Beshir, A.A.; Kuipers, F.A.; Van Mieghem, P.; Orda, A.

    In WDM networks, survivable routing and wavelength assignment (SRWA) involves assigning link-disjoint primary and backup lightpaths. In the on-line SRWA problem, a sequence of requests arrive and each request is either accepted or rejected based only on the input sequence seen so far. For special

  19. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    J.W. Davis


    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  20. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Curca-Tiving, F.; Opel, S.


    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  1. IOOS Data Portals and Uniform On-line Browse Capabilities

    Howard, M.; Currier, R. D.; Kobara, S.; Gayanilo, F.


    The Gulf of Mexico Coastal Ocean Observing System Regional Association (GCOOS-RA) is one of eleven Regional Associations organized under the NOAA-led U.S. Integrated Ocean Observing System (IOOS) Program Office. Each of the RAs operate standards-based regional data portals designed to aggregate near real-time and historical observed data and modeled outputs from distributed providers and to offer these and derived products in standardized ways to a diverse set of users. The RA's portals are based on the IOOS Data and Communications Plan which describes the functional elements needed for an interoperable system. One of these elements is called "Uniform On-line Browse" which is an informational service designed primarily to visualize the inventory of a portal. An on-line browse service supports the end user's need to discover what parameters are available, to learn the spatial and temporal extend of the holdings, and to examine the character of the data (e.g, variability, gappiness, etc). These pieces of information help the end user decide if the data are fit for his/her purpose and to construct valid data requests. Note that on-line browse is a distinctly different activity than data analysis because it seeks to yield knowledge about the inventory and not about what the data mean. "Uniform" on-line browse is a service that takes advantage of the standardization of the data portal's data access points. Most portals represent station locations on a map. This is a view of the data inventory but these plots are rarely generated by pulling data through the standards-based services offered to the end users but through methods only available to the portal programmers. This work will present results of Uniform On-line browse tools developed within GCOOS-RA and their applicability to other RA portals.

  2. Mentoring Narratives ON-LINE:Teaching the Principalship

    Allison I. Griffith


    Full Text Available The need to develop new models for preparation of school administrators has been a prominent concern in educational discourse in the last decade. Having been criticized for the inadequate preparation of the school leadership cadre, academic departments responsible for training future school administrators have had to revisit their approaches and to reframe their teaching philosophies to ensure the readiness of their graduates for the challenges and complexities of school leadership. This article reports on the new model of principals' training that has been used in York University's Principals' Qualification Program (PQP from the late 1990s onward. One component of the program brings traditional case methodology into a computer-mediated/on-line environment. The on-line cases are narratives from the everyday lives of the Ontario school administrators who serve as mentors in the on-line environment. Situating our discussion within the context of the rapidly changing educational landscape of Ontario, we focus on the PQP model to explore experientially generated case narratives as one method for teaching and learning the work of the local school administrator. We focus particularly on the teaching and learning embedded in computer-mediated or on-line case narratives used in training teachers for school leadership. We argue that the complexities of school leadership—the social, cultural, relational, ethical and moral context of school leadership—can be taught effectively through the reflective processes of on-line case narratives. We seek to contribute to the ongoing dialogue on the potential of new pedagogies and new technologies to help prepare the competent and responsible leaders for tomorrow's schools.


    Endiah Puji Hastuti


    Full Text Available Turbulensi aliran pendingin pada proses perpindahan panas berfungsi untuk meningkatkan nilai koefisien perpindahan panas, tidak terkecuali aliran dalam kanal bahan bakar. Program CFD (CFD=computational fluid dynamics, FLUENT adalah program komputasi berbasis elemen hingga (finite element yang mampu memprediksi dan menganalisis fenomena dinamika aliran fluida secara teliti. Program perhitungan CFD dipilih dalam penelitian ini karena selain akurat juga dapat memberikan visualisasi dengan baik. Penelitian ini bertujuan untuk memahami karakteristika perpindahan panas, massa dan momentum dari dinding rod bahan bakar ke pendingin secara visual, pada medan temperatur, medan tekanan, dan medan energi kinetika pendingin, sebagai fungsi dinamika aliran di dalam kanal, pada kondisi tunak dan transien. Analisis dinamika aliran pada kanal bahan bakar PWR berbasis CFD dilakukan dengan menggunakan sampel data reaktor PWR dengan daya 1000 MWe dengan susunan bahan bakar 17x17. Untuk menguji sensitivitas persamaan aliran yang sesuai dengan model aliran turbulen pada kanal bahan bakar dilakukan pemodelan dengan menggunakan persamaan k-omega (Ƙ-ω, k-epsilon (Ƙ-ε, dan Reynold stress model (RSM. Pada analisis sensitivitas aliran turbulen di dalam kanal digunakan model mesh hexahedral dengan memilih tiga geometri sel yang masing masing berukuran 0,5 mm; 0,2 mm dan 0,15 mm. Hasil analisis menunjukkan bahwa pada analisis kondisi tunak (steady state, terdapat hasil yang mirip pada model turbulen Ƙ-ε standard dan Ƙ-ω standard. Pengujian terhadap kriteria Dittus Boelter untuk bilangan Nusselt menunjukkan bahwa model Reynold stress model (RSM direkomendasikan. Analisis sensitivitas terhadap geometri mesh antara sel yang berukuran 0,5 mm, 0,2 mm dan 0,15 mm, menunjukkan bahwa geometri sel sebesar 0,5 mm telah mencukupi. Aliran turbulen berkembang penuh telah tercapai pada model LES dan DES, meskipun hanya dalam waktu singkat (3 s, model LES memerlukan waktu komputasi

  4. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Surian Pinem


    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  5. Comparative study of the contribution of various PWR spacer grid components to hydrodynamic and wall pressure characteristics

    Bhattacharjee, Saptarshi, E-mail: [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France); Ricciardi, Guillaume [Alternative Energies and Atomic Energy Commission (CEA) – Cadarache, DEN/DTN/STCP/LHC, 13108 Saint Paul lez Durance Cedex (France); Viazzo, Stéphane [Laboratoire de Mécanique, Modélisation et Procédés Propres (M2P2), UMR7340 CNRS, Aix-Marseille Université, Centrale Marseille, 13451 Marseille Cedex (France)


    Highlights: • Complex geometry inside a PWR fuel assembly is simulated using simplified 3D models. • Structured meshes are generated as far as possible. • Fluctuating hydrodynamic and wall pressure field are analyzed using LES. • Comparative studies between square spacer grid, circular spacer grid and mixing vanes are presented. • Simulations are compared with experimental data. - Abstract: Flow-induced vibrations in a pressurized water reactor (PWR) core can cause fretting wear in fuel rods. These vibrations can compromise safety of a nuclear reactor. So, it is necessary to know the random fluctuating forces acting on the rods which cause these vibrations. In this paper, simplified 3D models like square spacer grid, circular spacer grid and symmetric mixing vanes have been used inside an annular pipe. Hydrodynamic and wall pressure characteristics are evaluated using large eddy simulations (LES). Structured meshes are generated as far as possible. Simulations are compared with an experiment. Results show that the grid and vanes have a combined effect: grid accelerates the flow whereas the vanes contribute to the swirl structures. Spectral analysis of the simulations illustrate vortex shedding phenomenon in the wake of spacer grids. This initial study opens up interesting perspectives towards improving the modeling strategy and understanding the complex phenomenon inside a PWR core.

  6. Organ-specific gene expression in maize: The P-wr allele. Final report, August 15, 1993--August 14, 1996

    Peterson, T.A.


    The ultimate aim of our work is to understand how a regulatory gene produces a specific pattern of gene expression during plant development. Our model is the P-wr gene of maize, which produces a distinctive pattern of pigmentation of maize floral organs. We are investigating this system using a combination of classical genetic and molecular approaches. Mechanisms of organ-specific gene expression are a subject of intense research interest, as it is the operation of these mechanisms during eukaryotic development which determine the characteristics of each organism Allele-specific expression has been characterized in only a few other plant genes. In maize, organ-specific pigmentation regulated by the R, B, and Pl genes is achieved by differential transcription of functionally conserved protein coding sequences. Our studies point to a strikingly different mechanism of organ-specific gene expression, involving post-transcriptional regulation of the regulatory P gene. The novel pigmentation pattern of the P-wr allele is associated with differences in the encoded protein. Furthermore, the P-wr gene itself is present as a unique tandemly amplified structure, which may affect its transcriptional regulation.

  7. Astronomy On-Line Programme Enters "hot Week"


    World's Biggest Astronomy WWW-Event Attracts Thousands of Students The Astronomy On-line Programme (See ESO Press Release 09/96 of 18 June 1996) began officially on 1 October and is now about to enter its most intense phase, known as the Hot Week . On 18 - 22 November, an estimated 4000 astronomy-interested, mostly young people in Europe and on four other continents will get together during five days in what - not unexpected - has become the world's biggest astronomy event ever organised on the World Wide Web. This carefully structured Programme is carried out in collaboration between the European Association for Astronomy Education (EAAE), the European Southern Observatory and the European Commission, under the auspices of the Fourth European Week for Scientific and Technological Culture. The Programme has already had a most visible impact on the school education of natural sciences in various countries; for instance, the Internet-connection of schools has been advanced in some, in order to allow groups to participate. There have been numerous contacts among the groups across the borders and there are clear signs that many Astronomy On-line participants have progressed to use the impressive possibilities of the Web in an efficient and structured way. There has been a lively media interest in Astronomy On-line all over Europe and it is expected to increase during the next week. The current status of Astronomy On-line It is obvious that the pilot function of the Astronomy On-line Programme in the use of the Web has been very effective and that the associated dissemination of astronomical knowledge has been successful. At this time, more than 650 groups have registered with Astronomy On-line. Most come from 31 different European countries and a few dozen groups are located in North and South America as well as in Asia and Australia. Together they have experienced the steady build-up of Astronomy On-line over the past weeks, by means of numerous contributions from a

  8. Characterization of PWR vessel steel tearing under severe accident condition temperatures

    Matheron, Philippe, E-mail: [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Chapuliot, Stephane, E-mail: [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Nicolas, Laetitia, E-mail: [CEA, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Laboratoire de Mecanique des Structures Industrielles Durables, UMR CNRS-EDF 2832, 1 avenue du General de Gaulle, F-92141 Clamart (France); Koundy, Vincent, E-mail: [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Caroli, Cataldo, E-mail: [IRSN-DSR, Service d' evaluation des Accidents Graves et des Rejets radioactifs B.P. 17, 92262 Fontenay-aux-Roses Cedex (France)


    Highlights: Black-Right-Pointing-Pointer We characterized French PWR vessel steel tearing resistance at high temperatures. Black-Right-Pointing-Pointer Tearing tests on Compact Tension (CT) specimens were carried out. Black-Right-Pointing-Pointer The variability of tearing properties with PWR vessels specifications was studied. Black-Right-Pointing-Pointer We propose a tearing criterion (energy parameter Gfr) at high temperatures. - Abstract: In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead

  9. On-line Measuring Method for Shell Chamber Volume

    ZHANG Li-zhong; WANG De-min; JIANG Tao; CAO Guo-hua; WANG Qi


    Using the ideal gas state equation, an on-line measuring method for the shell chamber volume is studied in this paper. After analyzing how various measurement parameters affect the measurement accuracy, the system parameters are optimized in this method. Because the shape and volume of the tested items are similar, the method of using "tamping" to raise the accuracy and speed of the measurement is put forward. Based on the work above, a prototype of the testing instrument for shell chamber volume was developed, automatically testing and controlling. Compared with the method of "water weight", this method is more accurate, quicker and more automotive, so it is adaptable for the use of on-line detection.

  10. On-line Dynamic Security Assessment in Power Systems

    Weckesser, Johannes Tilman Gabriel

    tools may no longer be feasible, since they are generally based on extensive off-line studies. A core component of an efficient on-line dynamic security assessment is a fast and reliable contingency screening. As part of this thesis a contingency screening method is developed and its performance......The thesis concerns the development of tools and methods for on-line dynamic security assessment (DSA). In a future power system with low-dependence or even independence of fossil fuels, generation will be based to a large extent on noncontrollable renewable energy sources (RES), such as wind...... and solar radiation. Moreover, ongoing research suggests that demand response will be introduced to maintain power balance between generation and consumption at all times. Due to these changes the operating point of the power system will be less predictable and today’s stability and security assessment...

  11. On-line Ramsey Numbers for Paths and Stars

    Jaroslaw Grytczuk


    Full Text Available We study on-line version of size-Ramsey numbers of graphs defined via a game played between Builder and Painter: in one round Builder joins two vertices by an edge and Painter paints it red or blue. The goal of Builder is to force Painter to create a monochromatic copy of a fixed graph H in as few rounds as possible. The minimum number of rounds (assuming both players play perfectly is the on-line Ramsey number r(H of the graph H. We determine exact values of r(H for a few short paths and obtain a general upper bound r(Pn ≤ 4n-7. We also study asymmetric version of this parameter when one of the target graphs is a star Sn with n edges. We prove that r(Sn,H≤n ·e(H when H is any tree, cycle or clique.

  12. On-line corrosion monitoring in district heating systems

    Richter, Sonja; Thorarinsdottir, R.I.; Hilbert, Lisbeth Rischel


    complicates the chemistry of the environment. Hydrogen sulphide is present in geothermal systems and can be formed as a by-product of sulphate-reducing-bacteria (SRB). The application of electrochemical methods makes on-line monitoring possible. These methods include: Linear Polarization Resistance (LPR......), Electrochemical Noise (EN) and Zero Resistance Ammetry (ZRA). Electrochemical Resistance (ER) has also been used to measure corrosion. The method traditionally only measures corrosion off-line but with newly developed high-sensitive ER technique developed by MetriCorr in Denmark, on-line monitoring is possible......Traditionally corrosion monitoring in district heating systems has been performed offline via weight loss coupons. These measurements give information about the past and not the present situation and require long exposure time (weeks or months). The good quality of district heating medium makes...

  13. Fractal groups: Emergent dynamics in on-line learning communities

    Junia de Carvalho Fidelis Braga

    Full Text Available Drawing on complexity theory, this work discusses the complex dynamics and emergent patterns of on-line learning communities based on a doctoral study in the area of Applied Linguistics. The analysis will center on the interlocutions of fifty students who interacted in small groups without the teacher's direct intervention, in an undergraduate course offered by the School of Languages and Literature at the Federal University of Minas Gerais. By analyzing the data, I demonstrate that out of the interactions among the peers of autonomous on-line learning communities arise opportunities for the construction of shared meaning, distributed leadership, as well as other dynamics. I also demonstrate the fractal nature of these communities. Moreover, I discuss how these findings shed light on the creation and development of course designs for large groups.

  14. On-Line Metrology with Conoscopic Holography: Beyond Triangulation

    Álvarez, Ignacio; Enguita, Jose M.; Frade, María; Marina, Jorge; Ojea, Guillermo


    On-line non-contact surface inspection with high precision is still an open problem. Laser triangulation techniques are the most common solution for this kind of systems, but there exist fundamental limitations to their applicability when high precisions, long standoffs or large apertures are needed, and when there are difficult operating conditions. Other methods are, in general, not applicable in hostile environments or inadequate for on-line measurement. In this paper we review the latest research in Conoscopic Holography, an interferometric technique that has been applied successfully in this kind of applications, ranging from submicrometric roughness measurements, to long standoff sensors for surface defect detection in steel at high temperatures. PMID:22399984

  15. Escapist Motives for Playing On-Line Games

    Sudzina, Frantisek; Razmerita, Liana

    , Harteveld and Mayer’s framework (2009) of escapist motives, which identifies four main motives for playing on-line games: mundane breaking, stress relieving, pleasure seeking, and imagination conjuring. In the paper, we report preliminary findings from an exploratory questionnaire survey. Besides importance...... of escapist motives for playing Facebook and other on-line games, we investigate how they are linked to demographic data such as: age, gender, place of origin, along with other social interactions patterns and social network usage behavior, current gaming status and an estimate of gaming time. According......Social games have become popular along with the tremendous growth of social networking sites, esp. Facebook. There is a gap in literature on what motivates people to play Facebook games. This paper studies social games usage behavior of students. We focus on escapist reasons, based on Warmelink...

  16. On-line Corrosion Monitoring in District Heating Systems

    Richter, Sonja; Thorarinsdottir, R.I.; Hilbert, Lisbeth Rischel


    The corrosion control in district heating systems is today performed primarily with control of the water quality. The corrosion rate is kept low by assuring low dissolved oxygen concentration, high pH and low conductivity. Corrosion failures can occur, e.g. as a result of unknown oxygen ingress......, precipitation of deposits or crevices. The authors describe methods used for on-line monitoring of corrosion, cover the complications and the main results of a Nordic project....

  17. On-Line and Back at S.F.U.

    M. Sanderson


    Full Text Available Simon Fraser University library began operation with an automated circulation system. After deliberation, it mounted the first phase of a two-phase online circulation system. A radically revised loan pol·icy caused the system design and assumptions to be called into question. A cheaper, simpler, and more effective off-line system eventually replaced the on-line system. The systems, fiscal, and administrative implications of this decision are reviewed.

  18. Two Types of Designs for On-Line Circulation Systems

    Rob McGee


    Full Text Available On-line circulation systems divide into two types. One type contains records only for charged or otherwise absent items. The other contains a file of records for all titles or volumes in the library collection, regardless of their circulation status. This paper traces differences between the two types, examining different kinds of files and terminals, transaction evidence, the quality of bibliographic data, querying, and the possibility of functions outside circulation. Aspects of both operational and potential systems are considered.

  19. Differential Electronic Nose in On-Line Dynamic Measurements

    Osowski S.


    Full Text Available The paper presents application of differential electronic nose in the dynamic (on-line volatile measurement. First we compare the classical nose employing only one sensor array and its extension in the differential form containing two sensor arrays working in differential mode. We show that differential nose performs better at changing environmental conditions, especially the temperature, and well performs in the dynamic mode of operation. We show its application in recognition of different brands of tobacco

  20. On-line phase space measurement with kicker excitation

    Dietrich, J.; Maier, R.; Mohos, I.


    A new method for on-line phase space measurements with kicker excitation at COSY was developed. The position data were measured using the analog output of two beam position monitors (BPMs) and directly monitored on a digital storage oscilloscope with an external clock (bunch-synchronous sampling). Nonlinear behavior of the proton beam was visible as well as were resonance islands. Typical measurements are presented.

  1. The new on-line Czech Food Composition Database.

    Machackova, Marie; Holasova, Marie; Maskova, Eva


    The new on-line Czech Food Composition Database (FCDB) was launched on in December 2010 as a main freely available channel for dissemination of Czech food composition data. The application is based on a complied FCDB documented according to the EuroFIR standardised procedure for full value documentation and indexing of foods by the LanguaL™ Thesaurus. A content management system was implemented for administration of the website and performing data export (comma-separated values or EuroFIR XML transport package formats) by a compiler. Reference/s are provided for each published value with linking to available freely accessible on-line sources of data (e.g. full texts, EuroFIR Document Repository, on-line national FCDBs). LanguaL™ codes are displayed within each food record as searchable keywords of the database. A photo (or a photo gallery) is used as a visual descriptor of a food item. The application is searchable on foods, components, food groups, alphabet and a multi-field advanced search. Copyright © 2013 Elsevier Ltd. All rights reserved.

  2. Review of trigger and on-line processors at SLAC

    Lankford, A.J.


    The role of trigger and on-line processors in reducing data rates to manageable proportions in e/sup +/e/sup -/ physics experiments is defined not by high physics or background rates, but by the large event sizes of the general-purpose detectors employed. The rate of e/sup +/e/sup -/ annihilation is low, and backgrounds are not high; yet the number of physics processes which can be studied is vast and varied. This paper begins by briefly describing the role of trigger processors in the e/sup +/e/sup -/ context. The usual flow of the trigger decision process is illustrated with selected examples of SLAC trigger processing. The features are mentioned of triggering at the SLC and the trigger processing plans of the two SLC detectors: The Mark II and the SLD. The most common on-line processors at SLAC, the BADC, the SLAC Scanner Processor, the SLAC FASTBUS Controller, and the VAX CAMAC Channel, are discussed. Uses of the 168/E, 3081/E, and FASTBUS VAX processors are mentioned. The manner in which these processors are interfaced and the function they serve on line is described. Finally, the accelerator control system for the SLC is outlined. This paper is a survey in nature, and hence, relies heavily upon references to previous publications for detailed description of work mentioned here. 27 references, 9 figures, 1 table.


    J.D. Miller; C.L. Lin; G.H. Luttrell; G.T. Adel; Barbara Marin


    Washability analysis is the basis for nearly all coal preparation plant separations. Unfortunately, there are no on- line techniques for determining this most fundamental of all coal cleaning information. In light of recent successes at the University of Utah, it now appears possible to determine coal washability on-line through the use of x-ray computed tomography (CT) analysis. The successful development of such a device is critical to the establishment of process control and automated coal blending systems. In this regard, Virginia Tech, Terra Tek Inc., and U.S. coal producers have joined with the University of Utah and to undertake the development of an X-ray CT-based on- line coal washability analyzer with financial assistance from DOE. Each project participant brought special expertise to the project in order to create a new dimension in coal cleaning technology. The project involves development of appropriate software and extensive testing/evaluation of well-characterized coal samples from operating coal preparation plants. Data collected to date suggest that this new technology is capable of serving as a universal analyzer that can not only provide washability analysis, but also particle size distribution analysis, ash analysis, and perhaps pyritic sulfur analysis.


    J.D. Miller


    Washability analysis is the basis for nearly all coal preparation plant separations. Unfortunately, there are no on-line techniques for determining this most fundamental of all coal cleaning information. In light of recent successes at the University of Utah, it now appears possible to determine coal washability on-line through the use of x-ray computed tomography (CT) analysis. The successful development of such a device is critical to the establishment of process control and automated coal blending systems. In this regard, Virginia Tech, Terra Tek Inc., and several eastern coal companies have joined with the University of Utah and agreed to undertake the development of a x-ray CT-based on-line coal washability analyzer with financial assistance from DOE. The three-year project will cost $594,571, of which 33% ($194,575) will be cost-shared by the participants. The project involves development of appropriate software and extensive testing/evaluation of well-characterized coal samples from operating coal preparation plants. Each project participant brings special expertise to the project which is expected to create a new dimension in coal cleaning technology. Finally, it should be noted that the analyzer may prove to be a universal analyzer capable of providing not only washability analysis, but also particle size distribution analysis, ash analysis and perhaps pyritic sulfur analysis.

  5. Designing effective on-line continuing medical education.

    Zimitat, Craig


    The Internet, and new information and communication technologies available through the Internet, provides medical educators with an opportunity to develop unique on-line learning environments with real potential to improve physicians' knowledge and effect change in their clinical practice. There are approximately 100 websites offering on-line CME courses in the USA alone. However, few of these CME courses appear to be based on sound educational principles or CME research and may have little chance of achieving the broader goals of CME. The majority of these courses closely resemble their traditional counterparts (e.g. paper-based books are now electronic books) and appear to be mere substitutions for old-technology CME resources. Whilst some CME providers add unique features of the Internet to enrich their websites, they do not employ strategies to optimize the learning opportunities afforded by this new technology. The adoption of adult learning principles, reflective practice and problem-based approaches can be used as a foundation for sound CME course design. In addition, knowledge of Internet technology and the learning opportunities it affords, together with strategies to maintain participation and new assessment paradigms, are all needed for developing online CME. We argue for an evidence-based and strategic approach to the development of on-line CME courses designed to enhance physician learning and facilitate change in clinical behaviour.

  6. A note on on-line Ramsey numbers for quadrilaterals

    Joanna Cyman


    Full Text Available We consider on-line Ramsey numbers defined by a game played between two players, Builder and Painter. In each round Builder draws an the edge and Painter colors it either red or blue, as it appears. Builder's goal is to force Painter to create a monochromatic copy of a fixed graph \\(H\\ in as few rounds as possible. The minimum number of rounds (assuming both players play perfectly is the on-line Ramsey number \\(\\widetilde{r}(H\\ of the graph \\(H\\. An asymmetric version of the on-line Ramsey numbers \\(\\widetilde{r}(G,H\\ is defined accordingly. In 2005, Kurek and Ruciński computed \\(\\widetilde{r}(C_3\\. In this paper, we compute \\(\\widetilde{r}(C_4,C_k\\ for \\(3 \\le k \\le 7\\. Most of the results are based on computer algorithms but we obtain the exact value \\(\\widetilde{r}(C_4\\ and do so without the help of computer algorithms.

  7. Internet Teaching By Style: Profiling the On-line Professor

    Sharon Strand


    Full Text Available The goal of this article is to offer the results of a pilot study which examined the personality type and teaching style preferences of faculty who elected to teach an on-line course. The article will present a description of personality assessments, including the Myers-Briggs Type Indicator (MBTI and Anthony Gregorc's Transaction Ability Inventory used to determine teaching tendencies and styles. In addition, a structured written questionnaire developed by the researchers was used to assess teacher satisfaction with worldwide web-based instruction. Utilizing the results of these psychological assessments, a preliminary analysis of the personal characteristics of college professors who chose to teach on line will be presented. This pilot study found that some preferred teaching styles may be more compatible with the dynamics of distance learning formats. By determining successful teaching styles for on-line courses, we can develop more effective faculty development programs to assist others in successfully transitioning into the cyber-teaching and learning environment.

  8. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing


    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  9. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)


    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  10. VOF Calculations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    M. Murase


    Full Text Available We improved the computational grid and schemes in the VOF (volume of fluid method with the standard − turbulent model in our previous study to evaluate CCFL (countercurrent flow limitation characteristics in a full-scale PWR hot leg (750 mm diameter, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa. In this paper, therefore, to evaluate applicability of the VOF method to different fluid properties and a different scale, we did numerical simulations for full-scale air-water conditions and the 1/15-scale air-water tests (50 mm diameter, respectively. The results calculated for full-scale conditions agreed well with CCFL data and showed that CCFL characteristics in the Wallis diagram were mitigated under 1.5 MPa steam-water conditions comparing with air-water flows. However, the results calculated for the 1/15-scale air-water tests greatly underestimated the falling water flow rates in calculations with the standard − turbulent model, but agreed well with the CCFL data in calculations with a laminar flow model. This indicated that suitable calculation models and conditions should be selected to get good agreement with data for each scale.

  11. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail:; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)


    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  12. Test requirements for the integral effect test to simulate Korean PWR plants

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K


    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  13. Analysis of measured and calculated counterpart test data in PWR and VVER 1000 simulators

    d’Auria Francesco


    Full Text Available This paper presents an over view of the "scaling strategy", in particular the role played by the counter part test methodology. The recent studies dealing with a scaling analysis in light water reactor with special regard to the VVER 1000 Russian reactor type are presented to demonstrate the phenomena important for scaling. The adopted scaling approach is based on the selection of a few characteristic parameters chosen by taking into account their relevance in the behavior of the transient. The adopted computer code used is RELAP5/Mod3.3 and its accuracy has been demonstrated by qualitative and quantitative evaluation. Comparing experimental data, it was found that the investigated facilities showed similar behavior concerning the time trends, and that the same thermal hydraulic phenomena on a qualitative level could be predicted. The main results are: PSB and LOBI main parameters have similar trends. This fact is the confirmation of the validity of the adopted scaling approach and it shows that PWR and VVER reactor type behavior is very similar. No new phenomena occurred during the counter part test, despite the fact that the two facilities had a different lay out, and the already known phenomena were predicted correctly by the code. The code capability and accuracy are scale-independent. Both character is tics are necessary to permit the full scale calculation with the aim of nuclear power plant behavior prediction. .

  14. Replacement of Co-base alloy for radiation exposure reduction in the primary system of PWR

    Han, Jeong Ho; Nyo, Kye Ho; Lee, Deok Hyun; Lim, Deok Jae; Ahn, Jin Keun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, Sun Jin [Hanyang Univ., Seoul (Korea, Republic of)


    Of numerous Co-free alloys developed to replace Co-base stellite used in valve hardfacing material, two iron-base alloys of Armacor M and Tristelle 5183 and one nickel-base alloy of Nucalloy 488 were selected as candidate Co-free alloys, and Stellite 6 was also selected as a standard hardfacing material. These four alloys were welded on 316SS substrate using TIG welding method. The first corrosion test loop of KAERI simulating the water chemistry and operation condition of the primary system of PWR was designed and fabricated. Corrosion behaviors of the above four kinds of alloys were evaluated using this test loop under the condition of 300 deg C, 1500 psi. Microstructures of weldment of these alloys were observed to identify both matrix and secondary phase in each weldment. Hardnesses of weld deposit layer including HAZ and substrate were measured using micro-Vickers hardness tester. The status on the technology of Co-base alloy replacement in valve components was reviewed with respect to the classification of valves to be replaced, the development of Co-free alloys, the application of Co-free alloys and its experiences in foreign NPPs, and the Co reduction program in domestic NPPs and industries. 18 tabs., 20 figs., 22 refs. (Author).

  15. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail:, E-mail: [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)


    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  16. Study of chemical additives in the cementation of radioactive waste of PWR reactors

    Vieira, Vanessa Mota; Tello, Cledola Cassia Oliveira de, E-mail:, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors. It was cemented using two formulations, A and B, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The parameter evaluated in this study was the compressive strength at age of 28 days, is considered essential security issues relating to the handling, transport and storage of cemented waste product. The results showed that the addition of accelerators improved the compressive strength of the cemented products. (author)

  17. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)


    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  18. Numerical Simulation of Size Effects on Countercurrent Flow Limitation in PWR Hot Leg Models

    I. Kinoshita


    Full Text Available We have previously done numerical simulations using the two-fluid model implemented in the CFD software FLUENT6.3.26 to investigate effects of shape of a flow channel and its size on CCFL (countercurrent flow limitation characteristics in PWR hot leg models. We confirmed that CCFL characteristics in the hot leg could be well correlated with the Wallis parameters in the diameter range of 0.05 m≤D≤0.75 m. In the present study, we did numerical simulations using the two-fluid model for the air-water tests with D=0.0254 m to determine why CCFL characteristics for D=0.0254 m were severer compared with those in the range, 0.05 m≤D≤0.75 m. The predicted CCFL characteristics agreed with the data for D=0.0254 m and indicated that the CCFL difference between D=0.0254 m and 0.05 mm≤D≤0.75 mm was caused by the size effect and not by other factors.

  19. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.


    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  20. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)


    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  1. Regulatory Research of the PWR Severe Accident. Information Needs and Instrumentation for Hydrogen Control and Management

    Park, Gun Chul; Suh, Kune Y.; Lee, Jin Yong; Lee, Seung Dong [Seoul Nat' l Univ., Seoul (Korea, Republic of)


    The current research is concerned with generation of basic engineering data needed in the process of developing hydrogen control guidelines as part of accident management strategies for domestic nuclear power plants and formulating pertinent regulatory requirements. Major focus is placed on identification of information needs and instrumentation methods for hydrogen control and management in the primary system and in the containment, development of decision-making trees for hydrogen management and their quantification, the instrument availability under severe accident conditions, critical review of relevant hydrogen generation model and phenomena In relation to hydrogen behavior, we analyzed the severe accident related hydrogen generation in the UCN 3{center_dot}4 PWR with modified hydrogen generation model. On the basis of the hydrogen mixing experiment and related GASFLOW calculation, the necessity of 3-dimensional analysis of the hydrogen mixing was investigated. We examined the hydrogen control models related to the PAR(Passive Autocatalytic Recombiner) and performed MAAP4 calculation in relation to the decision tree to estimate the capability and the role of the PAR during a severe accident.

  2. PWR composite materials use. A particular case of safety-related service water pipes

    Pays, M.F.; Le Courtois, T


    This paper shows the present and future uses of composite materials in French nuclear and fossil-fuel power plants. Electricite de France has decided to install composite materials in service water piping in its future nuclear power plant (PWR) at Civaux (West of France) and for the firs time in France, in safety-related applications. A wide range of studies has been performed about the durability, the control and damage mechanisms of those materials under service conditions among an ongoing Research and Development project. The main results are presented under the following headlines: selection of basic materials and manufacturing processes; aging processes (mechanical behavior during `lifetime`); design rules; non destructive examination during manufacturing process and during operation. The studies have been focused on epoxy pipings. The importance of strong quality insurance policy requirements are outlined. A study of the use of composite pipes in power plants (hydraulic, fossil fuel, and nuclear) in France and around the world (USA, Japan, Western Europe) are presented whether it be safety related or non safety-related applications. The different technical solutions for materials and manufacturing processes are presented and an economic comparison is made between steel and composite pipes. (author) 2 refs.

  3. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi


    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  4. Evaluation of Fuel Performance Uncertainty in a PWR HFP RIA Analysis

    Lee, Joosuk; Woo, Swengwoong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)


    Sensitivity and combined uncertainty studies based on the various kinds of uncertainty sources have been carried out in a PWR hot full power (HFP) condition. - Cladding inner diameter, fuel thermal conductivity, fuel thermal expansion and peak power have induced a significant impact to the fuel enthalpy and temperature. - Cladding hoop strain was strongly affected by the uncertainty parameters of cladding inner diameter, fuel thermal expansion, EPRI-1 CHF and peak power. - Above results are valid in the given analysis condition in this paper. Thereby, the analysis conditions, for example the peak linear heat rate before RIA or peak power and FWHM etc, are changed the results will be changed also. Approved analysis methodology for licensing application in the safety analysis of reactivity initiated accident (RIA) in Korea is based on a conservative approach. But newly introduced safety criteria, described in section 4.2 of NUREG-0800, tend to reduce the margins or depending on the reactor types rod failure is predicted due to the pellet-to-cladding mechanical interaction (PCMI) criteria. Thereby, licensee is trying to improve the margins by utilizing a less conservative approach.

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

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


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

  6. Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

    Ge Shao


    Full Text Available To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.

  7. Irradiation Effects Test Series: Test IE-2. Test results report. [PWR

    Allison, C. M.; Croucher, D. W.; Ploger, S. A.; Mehner, A. S.


    The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. After one hour at 68 kW/m, a power-cooling-mismatch sequence was initiated by a flow reduction at constant power. At a flow of 2550 kg/s-m/sup 2/, the onset of film boiling occurred on one rod, Rod IE-011. An additional flow reduction to 2245 kg/s-m/sup 2/ caused the onset of film boiling on the remaining three rods. Data are presented on the behavior of fuel rods during quasiequilibrium and during film boiling operation. The effects of initial gap size, cladding irradiation, rod power cycling, a rapid power increase, and sustained film boiling are discussed. These discussions are based on measured test data, preliminary postirradiation examination results, and comparisons of results with FRAP-T3 computer model calculations.

  8. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  9. Studies of residual stress measurement and analysis techniques for a PWR dissimilar weld joint

    Ogawa, Naoki, E-mail: [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Muroya, Itaru; Iwamoto, Youichi; Ohta, Takahiro; Ochi, Mayumi; Hojo, Kiminobu [Mitsubishi Heavy Industries, Ltd., 2-1-1, Shinhama, Arai-cho, Takasago 676-8686 (Japan); Ogawa, Kazuo [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)


    For evaluation of the PWSCC crack propagation behavior, a test model was produced using the same fabrication process of Japanese PWR plants and the stress distribution change was measured during a fabrication process such as a hydrostatic test, welding a main coolant pipe to the stainless steel safe end and an operation condition test. For confirmation of validity of the numerical estimation method of the stress distribution, FE analysis was performed to calculate the stress distributions for each fabrication process. From the validation procedure, a standard residual stress evaluation method was established. Furthermore for consideration of characteristics of PWSCC's propagation behavior of the dissimilar welding joint of the safe end nozzles, the influence coefficients at the deepest point for the stress intensity factors of axial cracks with large aspect ratio a/c (crack depth/half of surface crack length) was prepared. The crack shape was assumed a rectangular shape and the stress intensity factors at the deepest point of the crack were calculated with change of crack depth using FE analysis. By using these stress distribution and influence coefficients, a behavior of a PWSCC crack propagation at the safe end nozzles can be estimated easily and rationally.

  10. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    Rouf; Su'ud, Zaki


    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  11. Performance of monosphere new gel type ion exchange resins for condensate polisher at PWR plants

    Nakanishi, S.; Nakamura, M.; Asou, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Izumi, T.; Deguchi, T.; Ino, T.; Hagiwara, M.


    There are two kinds of ion exchange resins of gel type and porous one which are used as condensate polisher in LWR nuclear power plants. In order to estimate the performance of these resins on the condensate polisher at the secondary cycle of Japanese PWR plants, a column test was performed setting the column test device in Ohi power station unit 1 of the Kansai Electric Power Co., Inc. and the variations of the resin properties and the samples at the end of column were analyzed. The column test showed that the cross-linking degree of the new gel resins used was lower than those of porous ones. The new resins captured larger amounts of Matrix-Diffused Crud than the conventional cation resins before regeneration but not after that. Whereas the surface adsorbed crud was less captured by the new resins than conventional anion resins. However, there were little differences among these resins in respects of rinsing characteristics, sphericity, water quality, break through capacity, etc. At the condensate polisher in the secondary system it was confirmed that new gel resins had almost the same performance as one of the conventional ones and could be applied to the actual plant. (M.N.)

  12. Fatigue Crack Growth Rate Behavior of Type 347 Stainless Steel in Simulated PWR Water Environment

    Min, Ki Deuk; Kim, Seon Jin [Hanyang University, Seoul (Korea, Republic of); Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    The pressurizer surge line of a Korean standard nuclear power plane uses Nb stabilized type 347 stainless steel. The pressurizer surge line is the pipe connecting the pressurizer and the hot leg line, and the path controlling the pressure and temperature of the cooling system of the nuclear reactor, operated at 316 .deg. C and in a 150atm. The pressurizer surge line operated at high temperature and high pressure receives thermal stress by a temperature change and mechanical stress by a pressure change at the same time, and by being exposed to the high temperature and high pressure cooling water environment of a nuclear power plant, environmental fatigue by stress and corrosion is the main damage instrument. As the effect of environmental fatigue has been reported, through low cycle fatigue, fatigue life evaluations of austenite stainless steel have been conducted, but evaluations of fatigue crack growth rate to evaluate the soundness are very poor. In this study, evaluated characteristics of fatigue crack growth rate base on a change of dissolved oxygen in a PWR environment

  13. MELCOR 1.8.2 assessment: Surry PWR TMLB` (with a DCH study)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.


    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified.

  14. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Brovchenko Mariya


    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  15. Development of a parametric containment event tree model of a severe PWR accident

    Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)


    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

  16. Development of neutron own codes for the simulation of PWR reactor core; Desarrollo de codigos neutronicos propios para la simulacion del nucleo de reactores PWR

    Ahnert, C.; Cabellos, O.; Garcia-Herranz, N.; Cuervo, D.; Herrero, J. J.; Jimenez, J.; Ochoa, R.


    The core physic simulation is enough complex to need computers and ad-hoc software, and its evolution is to best-estimate methodologies, in order to improve availability and safety margins in the power plant operation. the Nuclear Engineering Department (UPM) has developed the SEANAP System in use in several power plants in Spain, with simulation in 3D and at the pin level detail, of the nominal and actual core burnup, with the on-line surveillance, and operational maneuvers optimization. (Author) 8 refs.

  17. A Refreshable, On-line Cache for HST Data Retrieval

    Fraquelli, Dorothy A.; Ellis, Tracy A.; Ridgaway, Michael; DPAS Team


    We discuss upgrades to the HST Data Processing System, with an emphasis on the changes Hubble Space Telescope (HST) Archive users will experience. In particular, data are now held on-line (in a cache) removing the need to reprocess the data every time they are requested from the Archive. OTFR (on the fly reprocessing) has been replaced by a reprocessing system, which runs in the background. Data in the cache are automatically placed in the reprocessing queue when updated calibration reference files are received or when an improved calibration algorithm is installed. Data in the on-line cache are expected to be the most up to date version. These changes were phased in throughout 2015 for all active instruments.The on-line cache was populated instrument by instrument over the course of 2015. As data were placed in the cache, the flag that triggers OTFR was reset so that OTFR no longer runs on these data. "Hybrid" requests to the Archive are handled transparently, with data not yet in the cache provided via OTFR and the remaining data provided from the cache. Users do not need to make separate requests.Users of the MAST Portal will be able to download data from the cache immediately. For data not in the cache, the Portal will send the user to the standard "Retrieval Options Page," allowing the user to direct the Archive to process and deliver the data.The classic MAST Search and Retrieval interface has the same look and feel as previously. Minor changes, unrelated to the cache, have been made to the format of the Retrieval Options Page.

  18. On-line learning algorithms for locally recurrent neural networks.

    Campolucci, P; Uncini, A; Piazza, F; Rao, B D


    This paper focuses on on-line learning procedures for locally recurrent neural networks with emphasis on multilayer perceptron (MLP) with infinite impulse response (IIR) synapses and its variations which include generalized output and activation feedback multilayer networks (MLN's). We propose a new gradient-based procedure called recursive backpropagation (RBP) whose on-line version, causal recursive backpropagation (CRBP), presents some advantages with respect to the other on-line training methods. The new CRBP algorithm includes as particular cases backpropagation (BP), temporal backpropagation (TBP), backpropagation for sequences (BPS), Back-Tsoi algorithm among others, thereby providing a unifying view on gradient calculation techniques for recurrent networks with local feedback. The only learning method that has been proposed for locally recurrent networks with no architectural restriction is the one by Back and Tsoi. The proposed algorithm has better stability and higher speed of convergence with respect to the Back-Tsoi algorithm, which is supported by the theoretical development and confirmed by simulations. The computational complexity of the CRBP is comparable with that of the Back-Tsoi algorithm, e.g., less that a factor of 1.5 for usual architectures and parameter settings. The superior performance of the new algorithm, however, easily justifies this small increase in computational burden. In addition, the general paradigms of truncated BPTT and RTRL are applied to networks with local feedback and compared with the new CRBP method. The simulations show that CRBP exhibits similar performances and the detailed analysis of complexity reveals that CRBP is much simpler and easier to implement, e.g., CRBP is local in space and in time while RTRL is not local in space.

  19. On-line yields obtained with the ISOLDE RILIS

    Koester, U. E-mail:; Fedoseyev, V.N.; Andreyev, A.N.; Bergmann, U.C.; Catherall, R.; Cederkaell, J.; Dietrich, M.; De Witte, H.; Fedorov, D.V.; Fraile, L.; Franchoo, S.; Fynbo, H.; Georg, U.; Giles, T.; Gorska, M.; Hannawald, M.; Huyse, M.; Joinet, A.; Jonsson, O.C.; Kratz, K.L.; Kruglov, K.; Lau, Ch.; Lettry, J.; Mishin, V.I.; Oinonen, M.; Partes, K.; Peraejaervi, K.; Pfeiffer, B.; Ravn, H.L.; Seliverstov, M.D.; Thirolf, P.; Van de Vel, K.; Van Duppen, P.; Van Roosbroeck, J.; Weissman, L


    The ISOLDE resonance ionization laser ion source (RILIS) allows to ionize efficiently and selectively many metallic elements. In recent yield surveys and on-line experiments with the ISOLDE RILIS we observed {sup 23-34}Mg, {sup 26-34}Al, {sup 98-132}Cd, {sup 149}Tb, {sup 155-177}Yb, {sup 179-200}Tl, {sup 183-215}Pb and {sup 188-218}Bi. The obtained yields are presented together with measured release parameters which allow to extrapolate the release efficiency towards more exotic (short-lived) nuclides of the same elements.

  20. Integrated sensor array for on-line monitoring micro bioreactors

    Krommenhoek, Erik Eduard


    In this thesis the development of a microbioreactor array with integrated sensoss suitable for on-line screening of micro organisms is described. Therefore, an array of 2 micro bioreactors compatible with the 96-well microtiterplate format has been made and tested. The developed system was shown to 

  1. TELCAL: The On-line Calibration Software for ALMA

    Broguière, D.; Lucas, R.; Pardo, J.; Roche, J.-C.


    The ALMA on-line calibration regroups all the operations needed to maintain the ALMA interferometer optimally tuned to successfully execute the planned observations. The results of the calibrations are used in quasi-real time by the ALMA Control System. Since the first ALMA antennas were put into operation in 2009, TELCAL has been used for all the basic calibration operations and is still being improved following the project advancement. We describe here the calibrations done by TELCAL, its relationships with the other ALMA software subsystems and, briefly, the architecture of the software based on CORBA.

  2. Molecularly imprinted polymers for on-line extraction techniques.

    Moein, Mohammad M; Abdel-Rehim, Mohamed


    Recent years have seen an increasing interest in the use of molecularly imprinted polymers (MIPs) as a sorbent for different extraction methods and this is due to its high selectivity. The MIP is designed to show specificity for the analyte of interest. Moreover, MIPs show physical robustness, resistance to high temperatures and pressures, and stability in the presence of acids, bases and a wide range of organic solvents. In the present article, various novel sample preparation techniques which MIPs applied as sorbent and on-line connected with analytical instruments were highlighted and discussed. The future aspects of MIPs as well were described.

  3. Enhancing Learner Autonomy in an On-line Editing Programme

    Hebe Wong


    Full Text Available Little (1999 argues that in formal educational contexts, “the basis of learner autonomy is acceptance of responsibility for one’s own learning” (p.11. An autonomous learner takes responsibility for various aspects of learning (Benson & Voller, 1997; Holec, 1981. This study examines how learner autonomy opportunities were provided at various stages of writing in an on-line editing programme for a group of electronic engineering students and how the students took charge of their language learning when receiving feedback on their technical writing. The impact on their own learning effectiveness of the decisions students made is also discussed.

  4. On-line Adaptive Radiation Treatment of Prostate Cancer


    volume (GTV), mandible, rainstem, parotids , and lymph nodes. Another physician repeated he contouring on all planning and on-line images...computed entation. Note the consistency of delineation of gross olume; light blue, nodes; purple, parotid glands; green,k com ostregicomp urs on e...0.9 1.0 Mandible D S C in de x Patient # 1 2 3 4 5 6 7 0.5 0.6 0.7 0.8 0.9 1.0 Left Parotid D S C in de x Patient # Fig. 6. Dice similarity

  5. Adaptive calibration method with on-line growing complexity

    Šika Z.


    Full Text Available This paper describes a modified variant of a kinematical calibration algorithm. In the beginning, a brief review of the calibration algorithm and its simple modification are described. As the described calibration modification uses some ideas used by the Lolimot algorithm, the algorithm is described and explained. Main topic of this paper is a description of a synthesis of the Lolimot-based calibration that leads to an adaptive algorithm with an on-line growing complexity. The paper contains a comparison of simple examples results and a discussion. A note about future research topics is also included.

  6. On-line Measurements of Settling Charateristics in Activated Sludge

    Rasmussen, Michael R.; Larsen, Torben


    An on-line settling column for measuring the dynamic variations of settling velocity of activated sludge has been developed. The settling column is automatic and self-cleansing insuring continuous and reliable measurements. The settling column was tested on sludge from a batch reactor where sucrose...... was added as an impulse to activated sludge. The continuous measurement of settling velocity revealed a highly dynamic response after the sucrose was added. The result were verified with simultaneous measurement of the initial settling rate. A 200 hour experiment showed variations in settling velocity...

  7. Lower Bounds and Semi On-line Multiprocessor Scheduling

    T.C. Edwin Cheng


    Full Text Available We are given a set of identical machines and a sequence of jobs from which we know the sum of the job weights in advance. The jobs have to be assigned on-line to one of the machines and the objective is to minimize the makespan. An algorithm with performance ratio 1.6 and a lower bound of 1.5 is presented. This improves recent results by Azar and Regev who published an algorithm with performance ratio 1.625 for the less general problem that the optimal makespan is known in advance.

  8. On-line Monitoring and Active Control for Transformer Noise

    Liang, Jiabi; Zhao, Tong; Tian, Chun; Wang, Xia; He, Zhenhua; Duan, Lunfeng

    This paper introduces the system for on-line monitoring and active noise control towards the transformer noise based on LabVIEW and the hardware equipment including the hardware and software. For the hardware part, it is mainly focused on the composition and the role of hardware devices, as well as the mounting location in the active noise control experiment. And the software part introduces the software flow chats, the measurement and analysis module for the sound pressure level including A, B, C weighting methods, the 1/n octave spectrum and the power spectrum, active noise control module and noise data access module.

  9. Efficient and Secure Comparison for On-Line Auctions

    Damgård, Ivan Bjerre; Krøigaard, Mikkel; Geisler, Martin Joakim


    We propose a protocol for secure comparison of integers based on homomorphic encryption. We also propose a homomorphic encryption scheme that can be used in our protocol and makes it more efficient than previous solutions. Our protocol is well-suited for application in on-line auctions, both...... with respect to functionality and performance. It minimizes the amount of information bidders need to send, and for comparison of 16 bit numbers with security based on 1024 bit RSA (executed by two parties), our implementation takes 0.28 seconds including all computation and communication. Using precomputation...

  10. Robust Control Methods for On-Line Statistical Learning

    Capobianco Enrico


    Full Text Available The issue of controlling that data processing in an experiment results not affected by the presence of outliers is relevant for statistical control and learning studies. Learning schemes should thus be tested for their capacity of handling outliers in the observed training set so to achieve reliable estimates with respect to the crucial bias and variance aspects. We describe possible ways of endowing neural networks with statistically robust properties by defining feasible error criteria. It is convenient to cast neural nets in state space representations and apply both Kalman filter and stochastic approximation procedures in order to suggest statistically robustified solutions for on-line learning.

  11. On-line matrix addition for detecting aerosol particles

    ZHOU; Liuzhu; ZHU; Yuan; GUO; Xiaoyong; ZHAO; Wenwu; ZHENG; Haiyang; Gu; Xuejun; FANG; Li; ZHANG; Weijun


    Single aerosol particles were measured by matrix-assisted laser desorption/ionization (MALDI) with an aerosol time-of-flight mass spectrometer (ATOFMS). The inlet to the ATOFMS was coupled with an evaporation/condensation flow cell that allowed matrix addition by condensation onto the particles. The coated particles entered the ion source through three-stage differentially pumped capillary inlet and were then ionized by a focused 266 nm Nd:YAG laser. The mass spectra and aerodynamic size of the single particles can be obtained simultaneously. The on-line matrix addition technique makes it possible to identify biological aerosols in real-time.

  12. Analysis of the containment of a compact reactor PWR submitted to loss of coolant accident; Analise da contencao de um reator PWR compacto submetido a acidente de perda de refrigerante

    Dutra, Alexandre de Souza; Belchior Junior, Antonio; Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)


    In the present paper analyses were done with the computer code RELAP5/MOD2 for rising the process conditions of the containment of a compact reactor PWR of low potency, submitted to Loss of Coolant Accidents (LOCA). The main results obtained were the behavior of maximum conditions of pressure as a function of the available containment free volume. It was also studied the problem of containment sub-compartmentation, that is to say, the possibility of the rupture to happen in restricted spaces generating high sub-compartment peak pressure and, consequently, high strains on the internal structures. (author)

  13. The safety analysis and thermohydraulic methodologies for the power updating analyses in Spanish PWR plants; Methodologias de diseno termohidraulico y de analisis de seguridad en los aumentos de potencia de centrales PWR

    Salesa, F.


    This article describes the Safety Analysis and Thermohydraulic methodologies used by ENUSA for the Power Updating analyses in Spanish PWR plants of Westinghouse design: Design tools have been developed over the first cycles resulting new correlations of DNB, fitted to the new fuel assemblies, new DNBR calculation methodology and other improvements in the design areas. Using these methodologies, the available margins between design and limit values are wider. These new margins have allowed to accomplish the design criteria under the new power updating operational conditions. (Author)

  14. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.


    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  15. Optimization of the distribution of bars with gadolinium oxide in reactor fuel elements PWR; Optimizacion de la distribucion de barras con oxido de gadolinio en elementos combustibles para reactores PWR

    Melgar Santa Cecilia, P. A.; Velazquez, J.; Ahnert Iglesias, C.


    In the schemes of low leakage, currently used in the majority of PWR reactors, it makes use of absorbent consumables for the effective control of the factors of peak, the critical concentration of initial boron and the moderator temperature coefficient. One of the most used absorbing is the oxide of gadolinium, which is integrated within the fuel pickup. Occurs a process of optimization of fuel elements with oxide of gadolinium, which allows for a smaller number of configurations with a low peak factor for bar. (Author)

  16. Study for highly functional resin (macroporous resin) superior in removing micro particles in PWR primary circuit: on-site test

    Itou, A.; Kondo, K.; Kouzuma, Y., E-mail: [Kyusyu Electric Power Co., Inc., Minami-ku, Fukuoka (Japan); Umehara, R.; Shimizu, Y., E-mail: [Mitsubishi Heavy Industries, Ltd., Hyogo-ku, Kobe (Japan); Kogawa, N.; Nagamine, K., E-mail: [Nuclear Development Corp., Tokaimura, Ibaraki (Japan)


    In Japanese PWR plants, efforts to remove particulate constituents containing radioactive cobalt which provides a source of radiation exposure, are needed. Performance evaluation study was conducted for macroporous resin which was said to possess excellent performance in removing particulate constituents and whose practical accomplishment at plants in USA was reported to be good. As one of the means for radiation exposure reduction in PWR, a study for application of crud removing resin to actual plant was executed by laboratory experiments using simulated crud (Fe{sub 3}O{sub 4} particle). In this study, following two mechanisms were demonstrated as the particle capturing mechanism of macroporous resin; physical trapping by fine pores on resin surface; electrical adsorption onto resin surface. In addition, in parallel to the study for application of macroporous resin to actual PWR plant, on-site study was planned to investigate the primary system water chemistry during various stages of actual plant operation and to research performance of particle capturing in detail. As the on-site study, column experiments, there water was let pass through the column, were planned for various operation stage (startup period, power operation period and shutdown period). A kind of conventional gel-type resin and three kinds of macroporous resin were examined for onsite tests. As to particulate capturing, basic knowledge regarding capturing efficiency and influence of water chemistry on capturing performance were ordered. Capturing performance of each resin tested became clear and was ordered by comparison. Effectiveness of macroporous resin with regard to crud removal in primary coolant was confirmed. (author)

  17. Research of on-line detection system for power capacitor

    Yao, Junda; Qian, Zheng; Yu, Hao; Xia, Jiuyun


    The hidden danger exists in the power capacitor of power system due to long-time operation under the environment of high voltage. Thus, it is possible to induce serious fault, and the on-line detection system is urgently required. In this paper, two methods of the on-line detection system are compared in order to realize the better real-time condition detection. The first method is based on the STM microprocessor with an internal 12 bit A/D converter, which converts analog signals which is arrived from the sample circuit into digital signals, and then the FFT algorithm is used to accomplish the measurement of the voltage and current values of the capacitor. The second method is based on the special electric energy metering IC, which can obtain RMS (Root Mean Square) of voltage and current by processing the sampled data of the voltage and current, and store RMS of voltage and current in its certain registers. The operating condition of the capacitor can be obtained after getting the values of voltage and current. By comparing the measuring results of two methods, the second method could achieve a higher measurement accuracy and more simple construction.

  18. Polar On-Line Acquisition Relay and Transmission System (POLARATS)

    Yuracko, K.


    POLARATS (Polar On-Line Acquisition Relay And Transmission System) is being developed by YAHSGS LLC (YAHSGS) and Oak Ridge National Laboratory (ORNL) to provide remote, unattended monitoring of environmental parameters under harsh environmental conditions. In particular, instrumental design and engineering is oriented towards protection of human health in the Arctic, and with the additional goal of advancing Arctic education and research. POLARATS will obtain and transmit environmental data from hardened monitoring devices deployed in locations important to understanding atmospheric and aquatic pollutant migration as it is biomagnified in Arctic food chains. An Internet- and personal computer (PC)-based educational module will provide real time sensor data, on-line educational content, and will be integrated with workbooks and textbooks for use in middle and high school science programs. The educational elements of POLARATS include an Internet-based educational module that will instruct students in the use of the data and how those data fit into changing Arctic environments and food chains. POLARATS will: (1) Enable students, members of the community, and scientific researchers to monitor local environmental conditions in real time over the Internet; and (2) Provide additional educational benefits through integration with middle- and high-school science curricula. Information will be relayed from POLARATS devices to classrooms and libraries along with custom-designed POLARATS teaching materials that will be integrated into existing curricula to enhance the educational benefits realized from the information obtained.

  19. Increased Cortical Thickness in Professional On-Line Gamers

    Hyun, Gi Jung; Shin, Yong Wook; Kim, Bung-Nyun; Cheong, Jae Hoon; Jin, Seong Nam


    Objective The bulk of recent studies have tested whether video games change the brain in terms of activity and cortical volume. However, such studies are limited by several factors including cross-sectional comparisons, co-morbidity, and short-term follow-up periods. In the present study, we hypothesized that cognitive flexibility and the volume of brain cortex would be correlated with the career length of on-line pro-gamers. Methods High-resolution magnetic resonance scans were acquired in twenty-three pro-gamers recruited from StarCraft pro-game teams. We measured cortical thickness in each individual using FreeSurfer and the cortical thickness was correlated with the career length and the performance of the pro-gamers. Results Career length was positively correlated with cortical thickness in three brain regions: right superior frontal gyrus, right superior parietal gyrus, and right precentral gyrus. Additionally, increased cortical thickness in the prefrontal cortex was correlated with winning rates of the pro-game league. Increased cortical thickness in the prefrontal and parietal cortices was also associated with higher performance of Wisconsin Card Sorting Test. Conclusion Our results suggest that in individuals without pathologic conditions, regular, long-term playing of on-line games is associated with volume changes in the prefrontal and parietal cortices, which are associated with cognitive flexibility. PMID:24474988

  20. On-line estimation of concentration parameters in fermentation processes

    XIONG Zhi-hua; HUANG Guo-hong; SHAO Hui-he


    It has long been thought that bioprocess, with their inherent measurement difficulties and complex dynamics, posed almost insurmountable problems to engineers. A novel software sensor is proposed to make more effective use of those measurements that are already available, which enable improvement in fermentation process control. The proposed method is based on mixtures of Gaussian processes (GP) with expectation maximization (EM) algorithm employed for parameter estimation of mixture of models. The mixture model can alleviate computational complexity of GP and also accord with changes of operating condition in fermentation processes, i.e., it would certainly be able to examine what types of process-knowledge would be most relevant for local models' specific operating points of the process and then combine them into a global one. Demonstrated by on-line estimate of yeast concentration in fermentation industry as an example, it is shown that soft sensor based state estimation is a powerful technique for both enhancing automatic control performance of biological systems and implementing on-line monitoring and optimization.

  1. On-line structure-lossless digital mammogram image compression

    Wang, Jun; Huang, H. K.


    This paper proposes a novel on-line structure lossless compression method for digital mammograms during the film digitization process. The structure-lossless compression segments the breast and the background, compresses the former with a predictive lossless coding method and discards the latter. This compression scheme is carried out during the film digitization process and no additional time is required for the compression. Digital mammograms are compressed on-the-fly while they are created. During digitization, lines of scanned data are first acquired into a small temporary buffer in the scanner, then they are transferred to a large image buffer in an acquisition computer which is connected to the scanner. The compression process, running concurrently with the digitization process in the acquisition computer, constantly checks the image buffer and compresses any newly arrived data. Since compression is faster than digitization, data compression is completed as soon as digitization is finished. On-line compression during digitization does not increase overall digitizing time. Additionally, it reduces the mammogram image size by a factor of 3 to 9 with no loss of information. This algorithm has been implemented in a film digitizer. Statistics were obtained based on digitizing 46 mammograms at four sampling distances from 50 to 200 microns.

  2. On-Line Core Thermal-Hydraulic Model Improvement

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won


    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS.

  3. The Leuven isotope separator on-line laser ion source

    Kudryavtsev, Y; Franchoo, S; Huyse, M; Gentens, J; Kruglov, K; Müller, W F; Prasad, N V S; Raabe, R; Reusen, I; Van den Bergh, P; Van Duppen, P; Van Roosbroeck, J; Vermeeren, L; Weissman, L


    An element-selective laser ion source has been used to produce beams of exotic radioactive nuclei and to study their decay properties. The operational principle of the ion source is based on selective resonant laser ionization of nuclear reaction products thermalized and neutralized in a noble gas at high pressure. The ion source has been installed at the Leuven Isotope Separator On-Line (LISOL), which is coupled on-line to the cyclotron accelerator at Louvain-la-Neuve. sup 5 sup 4 sup , sup 5 sup 5 Ni and sup 5 sup 4 sup , sup 5 sup 5 Co isotopes were produced in light-ion-induced fusion reactions. Exotic nickel, cobalt and copper nuclei were produced in proton-induced fission of sup 2 sup 3 sup 8 U. The b decay of the sup 6 sup 8 sup - sup 7 sup 4 Ni, sup 6 sup 7 sup - sup 7 sup 0 Co, sup 7 sup 0 sup - sup 7 sup 5 Cu and sup 1 sup 1 sup 0 sup - sup 1 sup 1 sup 4 Rh isotopes has been studied by means of beta-gamma and gamma-gamma spectroscopy. Recently, the laser ion source has been used to produce neutron-d...

  4. Increased cortical thickness in professional on-line gamers.

    Hyun, Gi Jung; Shin, Yong Wook; Kim, Bung-Nyun; Cheong, Jae Hoon; Jin, Seong Nam; Han, Doug Hyun


    The bulk of recent studies have tested whether video games change the brain in terms of activity and cortical volume. However, such studies are limited by several factors including cross-sectional comparisons, co-morbidity, and short-term follow-up periods. In the present study, we hypothesized that cognitive flexibility and the volume of brain cortex would be correlated with the career length of on-line pro-gamers. High-resolution magnetic resonance scans were acquired in twenty-three pro-gamers recruited from StarCraft pro-game teams. We measured cortical thickness in each individual using FreeSurfer and the cortical thickness was correlated with the career length and the performance of the pro-gamers. CAREER LENGTH WAS POSITIVELY CORRELATED WITH CORTICAL THICKNESS IN THREE BRAIN REGIONS: right superior frontal gyrus, right superior parietal gyrus, and right precentral gyrus. Additionally, increased cortical thickness in the prefrontal cortex was correlated with winning rates of the pro-game league. Increased cortical thickness in the prefrontal and parietal cortices was also associated with higher performance of Wisconsin Card Sorting Test. Our results suggest that in individuals without pathologic conditions, regular, long-term playing of on-line games is associated with volume changes in the prefrontal and parietal cortices, which are associated with cognitive flexibility.

  5. Precise On-line Position Measurement for Particle Therapy

    Actis, O; König, S


    An on-line beam position monitoring and regular beam stability tests are of utmost importance for the Quality Assurance (QA) of the patient treatment at any particle therapy facility. The Gantry${0.5 mm}2$ at the Paul Scherrer Institute uses a strip ionization chamber for the on-line beam position verification. The design of the strip chamber placed in the beam in front of the patient allows for a small beam penumbra in order to achieve a high-quality lateral beam delivery. The detector granularity and the low noise allow the reconstruction of the signals offered by Gantry${0.5 mm}2$ with a precision of about 0.1 mm. The frond-end electronics and the whole data processing sequence have been optimized for minimizing the dead time between the beam applications to about 2 ms: the charge collection is performed in about 1 ms, read-out takes place in 100 $\\mu$s while data verification and logging are completed in less than 1 ms. The sub-millimeter precision of the lateral reconstruction allows the dose inhomogenei...

  6. Fully On-line Introductory Physics with a Lab

    Schatz, Michael

    We describe the development and implementation of a college-level introductory physics (mechanics) course and laboratory that is suited for both on-campus and on-line environments. The course emphasizes a ``Your World is Your Lab'' approach whereby students first examine and capture on video (using cellphones) motion in their immediate surroundings, and then use free, open-source software both to extract data from the video and to apply physics principles to build models that describe, predict, and visualize the observations. Each student reports findings by creating a video lab report and posting it online; these video lab reports are then distributed to the rest of the class for peer review. In this talk, we will discuss the student and instructor experiences in courses offered to three distinct audiences in different venues: (1) a Massively Open On-line Course (MOOC) for off-campus participants, (2) a flipped/blended course for on-campus students, and, most recently, (3) a fully-online course for off-campus students.

  7. Containment fan cooler heat transfer calculation during main steam line break for Maanshan PWR plant

    Yuann, Yng-Ruey, E-mail:; Kao, Lain-Su, E-mail:


    Highlights: • Evaluate component cooling water (CCW) thermal response during MSLB for Maanshan. • Using GOTHIC to calculate CCW temperature and determine time required to boil CCW. • Both convective and condensation heat transfer from the air side are considered. • Boiling will not occur since T{sub B} is sufficiently longer than CCW pump restart time. -- Abstract: A thermal analysis has been performed for the Containment Fan Cooler Unit (FCU) during Main Steam Line Break (MSLB) accident, concurrent with loss of offsite power, for Maanshan PWR plant. The analysis is performed in order to address the waterhammer and two-phase flow issues discussed in USNRC's Generic Letter 96-06 (GL 96-06). Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt for each unit. The design basis for containment temperature is Main Steam Line Break (MSLB) accident at power of 2830.5 MWt, which results in peak vapor temperature of 387.6 °F. The design is such that when MSLB occurs concurrent with loss of offsite power (MSLB/LOOP), both the coolant pump on the secondary side and the fan on the air side of the FCU loose power and coast down. The pump has little inertia and coasts down in 2–3 s, while the FCU fan coasts down over much longer period. Before the pump is restored through emergency diesel generator, there is potential for boiling the coolant in the cooling coils by the high-temperature air/steam mixture entering the FCU. The time to boiling depends on the operating pressure of the coolant before the pump is restored. The prediction of the time to boiling is important because it determines whether there is potential for waterhammer or two-phase flow to occur before the pump is restored. If boiling occurs then there exists steam region in the pipe, which may cause the so called condensation induced waterhammer or column closure waterhammer. In either case, a great amount of effort has to be spent to


    Tagor Malem Sembiring


    Full Text Available Setelah kejadian Fukushima, penggunaan sistem keselamatan pasif menjadi persyaratan yang penting untuk PLTN. PLTN jenis PWR maju kelas 1000 yang didesain oleh Westinghouse, AP1000, memiliki fitur keselamatan pasif disamping sederhana dan modular. Sebelum memilih suatu PLTN, maka perlu dilakukan suatu evaluasi terhadap parameter desainnya. Salah satu parameter yang penting dalam keselamatan adalah kritikalitas teras. Permasalahan pokok dalam mengevaluasi parameter kritikalitas teras AP1000 tidak adanya data komposisi material SS304 dan H2O di daerah reflektor dan diameter penyerap SS304. Dengan demikian tujuan penelitian ini adalah mendapatkan model teras 3-dimensi AP1000 dan siap diaplikasikan dalam evaluasi parameter kritikalitas teras. Hasil perhitungan menunjukkan bahwa komposisi terbaik SS304 dan H2O di reflektor teras bagian atas dan bawah masing-masing 50 vol%, sedangkan diameter penyerap SS304 adalah 0,960 cm. Evaluasi konsentrasi boron kritis menunjukkan perbedaan yang signifikan dengan nilai desain. Meskipun penyebab utama dari perbedaan ini belum diketahui, akan tetapi dapat dibuktikan bahwa konsentrasi boron kritis sangat sensitif dengan densitas UO2. Untuk reaktivitas padam, reaktor AP1000 memiliki margin subkritikalitas teras yang besar untuk satu siklus operasi. Dengan demikian teras yang diusulkan dapat digunakan sebagai acuan untuk evaluasi parameter teras lainnya atau perangkat analitis lainnya dalam rangka mengevaluasi desain reaktor AP1000. Kata kunci: AP1000, kritikalitas, konsentrasi boron kritis, reaktivitas padam   After the Fukushima accident, the use of passive safety system becomes an important requirement for the nuclear power plant (NPP. The advanced PWR NPP with 1000 MW (electric class, designed by Westinghouse, AP1000, a reactor with the passive safety features as well as simple and modular. Before selecting a nuclear power plant, there should be an evaluation of the design parameter. One important parameter in

  9. Valve inlet fluid conditions for pressurizer safety and relief valves for B and W 177-FA and 205-FA plants. Final report. [PWR

    Cartin, L.R.; Winks, R.W.; Merchent, J.W.; Brandt, R.T.


    The overpressurization transients for the Babcock and Wilcox Company's 177- and 205-FA units are reviewed to determine the range of fluid conditions expected at the inlet of pressurizer safety and relief valves. The final Safety Analysis Report, extended high-pressure injection, and cold overpressurization events are considered. The results of this review, presented in the form of tables and graphs, provide input to the PWR utilities in their justification that the fluid conditions under which their valve designs were tested as part of the EPRI PWR Safety and Relief Valve Test Program are representative of those expected in their unit(s).

  10. A MATLAB-Linked Solver to Find Fuel Depletion in a PWR, a Suggested VVER-1000 Type

    F. Faghihi


    Full Text Available Coupled first-order IVPs are frequently used in many parts of engineering and sciences. We present a “solver” including three computer programs which were joint with the MATLAB software to solve and plot solutions of the first-order coupled stiff or nonstiff IVPs. Some applications related to IVPs are given here using our MATLAB-linked solver. Muon catalyzed fusion in a D-T mixture is considered as a first dynamical example of the coupled IVPs. Then, we have focused on the fuel depletion in a suggested PWR including poisons burnups (xenon-135 and samarium-149, plutonium isotopes production, and uranium depletion.

  11. Approaches to analyze the bowing of German PWR fuel assemblies; Ansaetze zur Analyse des Biegeverhaltens deutscher DWR-Brennelemente

    Boeke, H.; Bauer, R.; Bloemeling, F.; Lawall, R. [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)


    The analysis of the bowing behavior of PWR fuel elements is required in case of increased fuel element deformations that have been observed during the last years. In the contribution the following issues are discussed: fuel element properties (stiffness, constructive features), influence factors (guiding tubes, spacer), load transfer and its impact. Under consideration of external boundary conditions an evaluation scheme was developed, using analysis data (control rod drop time), friction force measurements, fuel element characteristics (fuel element deformation, bowing) and their ranking, and simulation models (fluid-structure interactions). The evaluation scheme allows the definition of appropriate measures. The suitability of the methodology was demonstrated.

  12. Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

    Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan); Mochizuki, Hiroki [The Japan Research Institute Ltd., Tokyo (Japan)


    For providing conservative PWR spent fuel compositions from the view point of nuclear criticality safety, correction factors applicable for result of burnup calculation by ORIGEN2 were evaluated. Its conservativeness was verified by criticality calculations using MVP. To calculate these correction factors, analyses of spent fuel isotopic composition data were performed by ORIGEN2. Maximum or minimum value of the ratio of calculation result to experimental data was chosen as correction factor. These factors are given to each set of fuel assembly and ORIGEN2 library. They could be considered as the re-definition of recommended isotopic composition given in Nuclear Criticality Safety Handbook. (author)

  13. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)


    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  14. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Kim, Byung Cheol; Chang, Ki Oak


    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  15. VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

    Salko, Robert K [ORNL; Sung, Yixing [Westinghouse Electric Company, Cranberry Township; Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township; Xu, Yiban [Westinghouse Electric Company, Cranberry Township; Cao, Liping [Westinghouse Electric Company, Cranberry Township


    The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time step of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.

  16. PWR core and spent fuel pool analysis using scale and nestle

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)


    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  17. Burn-up credit in criticality safety of PWR spent fuel

    Mahmoud, Rowayda F., E-mail: [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)


    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  18. Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

    Ferroukhi, H.; Vasiliev, A. [Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Dufresne, A. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Chawla, R. [Dept. of Physics, EPFL, 1015 Lausanne (Switzerland); Paul Scherrer Institut (Switzerland)


    Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

  19. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.


    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  20. Performance evaluation of PSO and GA in PWR core loading pattern optimization

    Khoshahval, F., E-mail: [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Minuchehr, H. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Zolfaghari, A., E-mail: [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of)


    Research highlights: The performance of both GA and PSO methods in optimizing of a PWR core are adequate. It seems GA arrives to its final parameter value in a fewer generation than the PSO. The computation time for GA is higher than PSO. The GA-2 and PSO-CFA algorithms perform better in comparison to GA-1 and PSO-IWA. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and species evolution. GA is finding popularity as design tools because of its versatility, intuitiveness and ability to solve highly non-linear, mixed integer optimization problems. PSO refers to a relatively new family of algorithms and is mainly inspired by social behavior patterns of organisms that live within large group. This study addresses the application and performance comparison of PSO and GA optimization methods for nuclear fuel loading pattern problem. Flattening of power inside the reactor core of Bushehr nuclear power plant (WWER-1000 type) is chosen as an objective function to prove the validity of algorithms. In addition the performance of both optimization techniques in terms of convergence rate and computational time is compared. It is found that, from an evolutionary point of view, the performance of both GA and PSO is quite adequate. But, GA seems to arrive at its final parameter value in a fewer generations than the PSO. It is also noticed that, the computation time for implemented GA in this work is too high in comparison to PSO.

  1. Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

    Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier


    As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard

  2. Severe accident modeling of a PWR core with different cladding materials

    Johnson, S. C. [Westinghouse Electric Company LLC, 5801 Bluff Road, Columbia, SC 29209 (United States); Henry, R. E.; Paik, C. Y. [Fauske and Associates, Inc., 16W070 83rd Street, Burr Ridge, IL 60527 (United States)


    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  3. FEDIX on-line information service: Design, develop, test, and implement, an on-line research and education information service

    Rodman, J.A.


    The FEDIX Annual Status Report provides details regarding an on-line information project designed, developed and implemented by Federal Information Exchange, Inc., a diversified information services company. This document details the project design activities, summarizes the developmental phases of the project and describes the implementation activities generated to fulfill the project's objectives. The information contained in this document illustrates FIE's continuing commitment to serve as the link that facilitates the dissemination of federal information to the education community. This report reviews the project accomplishments and describes intended service enhancements.

  4. Assessment of PWR fuel degradation by post-irradiation examinations and modeling in DEGRAD-1 code; Avaliacao da degradacao de combustivel PWR por exames pos-irradiacao e modelagem no codigo DEGRAD-1

    Castanheira, Myrthes; Lucki, Georgi; Silva, Jose Eduardo Rosa da; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear]. E-mail: myrthes@ipen


    On the majority of the cases, the inquiries on primary failures and secondary in PWR fuel rods are based on results of analysis were made use of the non-destructive examination results (coolant activities monitoring, sipping tests, visual examination). The complementary analysis methodology proposed in this work includes a modeling approach to characterization of the physical effects of the individual chemistry mechanisms that constitute the incubation phase of degradation phenomenon after primary failure that are integrated in the reactor operational history under stationary operational regime, and normal power transients. The computational program called DEGRAD-1 was developed based on this modeling approach. The practical outcome of the program is to predict cladding regions susceptible to massive hydriding. The applications presented demonstrate the validity of proposed method and models by actual cases simulation, which (primary and secondary) defects positions were known and formation time was estimated. By using the modeling approach, a relationship between the hydrogen concentration in the gap and the inner cladding oxide thickness has been identified which, when satisfied, will induce massive hydriding. The novelty in this work is the integrated methodology, which supplements the traditional analysis methods (using data from non-destructive techniques) with mathematical models for the hydrogen evolution, oxidation and hydriding that include refined approaches and criteria for PWR fuel, and using the FRAPCON-3 fuel performance code as the basic tool. (author)

  5. Configuration Database for BaBar On-line

    Salnikov, Andrei


    The configuration database is one of the vital systems in the BaBar on-line system. It provides services for the different parts of the data acquisition system and control system, which require run-time parameters. The original design and implementation of the configuration database played a significant role in the successful BaBar operations since the beginning of experiment. Recent additions to the design of the configuration database provide better means for the management of data and add new tools to simplify main configuration tasks. We describe the design of the configuration database, its implementation with the Objectivity/DB object-oriented database, and our experience collected during the years of operation.

  6. On-line corrosion monitoring in district heating systems

    Richter, Sonja; Thorarinsdottir, R.I.; Hilbert, Lisbeth Rischel


    Traditionally corrosion monitoring in district heating systems has been performed offline via weight loss coupons. These measurements give information about the past and not the present situation and require long exposure time (weeks or months). The good quality of district heating medium makes...... corrosion monitoring a challenge. Under normal conditions the pH is high (app. 9), conductivity is low (app. 10-200 µS/cm) and the concentration of dissolved oxygen is negligible. The low corrosion rates (in the order of µm/y) are difficult to measure and furthermore, factors such as hydrogen sulphide......), Electrochemical Noise (EN) and Zero Resistance Ammetry (ZRA). Electrochemical Resistance (ER) has also been used to measure corrosion. The method traditionally only measures corrosion off-line but with newly developed high-sensitive ER technique developed by MetriCorr in Denmark, on-line monitoring is possible...

  7. The Task Manager for the LHCb On-Line Farm

    Bonifazi, F; Carbone, A; Galli, D; Gregori, D; Marconi, U; Peco, G; Vagnoni, V


    The Task Manager is a utility to start, stop and list processes on the on-line farm. Each process started by the Task Manager has a string environment variable set, named UTGID (User defined unique Thread Group Identifier) which allows to identify the process. The Task Manager uses the UTGID to list the running processes and to identify the processes to be stopped. It has also the ability to start a process using a particular user name and to set the scheduler type and the priority for the process itself. The Task Manager package includes a Linux DIM server (tmSrv), four Linux command line DIM clients (tmStart, tmLs, tmKill and tmStop) and a JCOP (Joint Control Project) PVSS client.

  8. On-line Test for Train Communication Based System

    Zeng Xiaoqing; Masayuki Matsumoto; Kinji Mori; XU Fucang


    This paper gives out a new train automatic control system, which is based on train communication, and proposes a high assurance method to construct the system from current system. In current automatic train control (ATC) system, the central logic device detects position of each train and calculates permissible speed of each blocking section. Therefore, the central logic device controls speed of all trains. On the contrary, in the new system proposed in this paper, there is no central logical device and, train can communicate each other. The train detects the position and calculates the permissible speed itself according to the received position information of the preceding train. In the traditional method of changing an old system to a new one, test must be done off-line.While the integration technique proposed in this paper achieves on-line properties, and high assurance can be satisfied.

  9. On-line methods for rotorcraft aeroelastic mode identification

    Molusis, J. A.; Kleinman, D. L.


    The requirements for the on-line identification of rotorcraft aeroelastic blade modes from random response test data are presented. A recursive maximum likelihood (RML) technique is used in conjunction with a bandpass filter to identify isolated blade mode damping and frequency. The RML technique is demonstrated to have excellent convergence characteristics in random measurement noise and random process noise excitation. The RML identification technique uses an ARMA representation for the aeroelastic stochastic system and requires virtually no user interaction while providing accurate confidence bands on the parameter estimates. Comparisons are made with an off-line Newton type maximum likelihood algorithm which uses a state variable model representation. Results are presented from simulation random response data which quantify the identifed parameter convergence behavior for various levels of random excitation which is typical of wind tunnel turbulence levels. The RML technique is applied to hingless rotor test data from the NASA Langley Research Center Helicopter Hover Facility.

  10. An on-line method in studying music parsing.

    Berent, I; Perfetti, C A


    Listening to music entails the construction of a mental representation based on partial and ambiguous information. This study examines an experimental method that reflects such parsing decisions on-line by detecting the cognitive load resulting from temporary parsing failures. The method investigated was a divided attention paradigm in which listening to music was the primary task and click detection was a concurrent secondary task. It was hypothesized that increasing the complexity of the primary task by introducing an unprepared chromatic modulation results in an increase in response latencies to a click presented immediately after the modulatory shift. The support of this prediction by musicians' data provides evidence for the sensitivity of the paradigm. The failure of non-musicians to reflect the expected load is attributed to their attention-allocation strategy. These results are discussed in terms of their implications on the view of the musical parser as deterministic.

  11. 3rd Computer Science On-line Conference

    Senkerik, Roman; Oplatkova, Zuzana; Silhavy, Petr; Prokopova, Zdenka


    This book is based on the research papers presented in the 3rd Computer Science On-line Conference 2014 (CSOC 2014).   The conference is intended to provide an international forum for discussions on the latest high-quality research results in all areas related to Computer Science. The topics addressed are the theoretical aspects and applications of Artificial Intelligences, Computer Science, Informatics and Software Engineering.   The authors provide new approaches and methods to real-world problems, and in particular, exploratory research that describes novel approaches in their field. Particular emphasis is laid on modern trends in selected fields of interest. New algorithms or methods in a variety of fields are also presented.   This book is divided into three sections and covers topics including Artificial Intelligence, Computer Science and Software Engineering. Each section consists of new theoretical contributions and applications which can be used for the further development of knowledge of everybod...

  12. On-Line Adaptive Radiation Therapy: Feasibility and Clinical Study

    Taoran Li


    Full Text Available The purpose of this paper is to evaluate the feasibility and clinical dosimetric benefit of an on-line, that is, with the patient in the treatment position, Adaptive Radiation Therapy (ART system for prostate cancer treatment based on daily cone-beam CT imaging and fast volumetric reoptimization of treatment plans. A fast intensity-modulated radiotherapy (IMRT plan reoptimization algorithm is implemented and evaluated with clinical cases. The quality of these adapted plans is compared to the corresponding new plans generated by an experienced planner using a commercial treatment planning system and also evaluated by an in-house developed tool estimating achievable dose-volume histograms (DVHs based on a database of existing treatment plans. In addition, a clinical implementation scheme for ART is designed and evaluated using clinical cases for its dosimetric qualities and efficiency.

  13. The Monitor System for the LHCb on-line farm

    Bonifazi, F; Carbone, A; Galli, D; Gregori, D; Marconi, U; Peco, G; Vagnoni, V


    The aim of the LHCb on-line farm Monitor System is to keep under control all the working indicators which are relevant for the farm operation, and to set the appropriate alarms whenever an error or a critical condition comes up. Since the most stressing tasks of the farm are the data transfer and processing, relevant indicators includes the CPU and the memory load of the system, the network interface and the TCP/IP stack parameters, the rates of the interrupts raised by the network interface card and the detailed status of the running processes. The monitoring of computers’ physical conditions (temperatures, fan speeds and motherboard voltages) are the subject of a separate technical note, since they are accessed in a different way, by using the IPMI protocol.

  14. The Message Logger for the LHCb on-line farm

    Bonifazi, F; Carbone, A; Galli, D; Gregori, D; Marconi, U; Peco, G; Vagnoni, V


    The Message Logger is a utility which provide a logger facility for the processes running on the nodes of the on-line farm. It can also be used to collect the processes stdout/stderr. The Message Logger can be exploited using two different policies: either as a no-drop logger facility (messages cannot be lost, but a write to the logger facility blocks in case of full-buffer condition, due e.g. to a network congestion) or as a congestion-proof logger facility (a write to the logger facility never locks even in case of network congestion, but, in this case, messages are dropped). The Message Logger package includes a Linux DIM server (logSrv), a Linux terminal/command-line DIM client (logViewer) and a PVSS DIM client.

  15. On-line blind separation of non-stationary signals

    Todorović-Zarkula Slavica


    Full Text Available This paper addresses the problem of blind separation of non-stationary signals. We introduce an on-line separating algorithm for estimation of independent source signals using the assumption of non-stationary of sources. As a separating model, we apply a self-organizing neural network with lateral connections, and define a contrast function based on correlation of the network outputs. A separating algorithm for adaptation of the network weights is derived using the state-space model of the network dynamics, and the extended Kalman filter. Simulation results obtained in blind separation of artificial and real-world signals from their artificial mixtures have shown that separating algorithm based on the extended Kalman filter outperforms stochastic gradient based algorithm both in convergence speed and estimation accuracy.

  16. A Fingerprint Minutiae Matching Method Based on Line Segment Vector


    Minutiae-based fingerprint matching is the most commonly used in an automatic fingerprint identification system. In this paper, we propose a minutia matching method based on line segment vector. This method uses all the detected minutiae (the ridge ending and the ridge bifurcation) in a fingerprint image to create a set of new vectors (line segment vector). Using these vectors, we can determine a truer reference point more efficiently. In addition, this new minutiae vector can also increase the accuracy of the minutiae matching. By experiment on the public domain collections of fingerprint images fvc2004 DB3 set A and DB4 set A, the result shows that our algorithm can obtain an improved verification performance.

  17. A versatile apparatus for on-line emission channeling experiments

    da Silva, Manuel Ribeiro; Correia, João Guilherme; Amorim, Lígia Marina; Pereira, Lino Miguel da Costa


    The concept and functionality of an apparatus dedicated to emission channeling experiments using short-lived isotopes on-line at ISOLDE/CERN is described. The setup is assembled in two functional blocks - (a) base stand including beam collimation, implantation and measurement chamber, cryogenic extension, and vacuum control system and - (b) Panmure goniometer extension including maneuvering cradle and sample heating furnace. This setup allows for in situ implantation and sample analysis in the as-implanted state and upon cooling down to 50 K and during annealing up to 1200 K. The functionality of the setup will be illustrated with the example of establishing the lattice location of $^{56}$Mn probes implanted into GaAs.

  18. On-line control of the nonlinear dynamics for synchrotrons

    Bengtsson, J.; Martin, I. P. S.; Rowland, J. H.; Bartolini, R.


    We propose a simple approach to the on-line control of the nonlinear dynamics in storage rings, based on compensation of the nonlinear resonance driving terms using beam losses as the main indicator of the strength of a resonance. The correction scheme is built on the analysis of the resonance driving terms in first perturbative order and on the possibility of using independent power supplies in the sextupole magnets, which is nowadays present in many synchrotron light sources. Such freedom allows the definition of "smart sextupole knobs" attacking each resonance separately. The compensation scheme has been tested at the Diamond light source and proved to be effective in opening up the betatron tune space, resonance free, available to the electron beam and to improve the beam lifetime.

  19. BCal: an on-line Bayesian radiocarbon calibration tool

    Caitlin E. Buck


    Full Text Available In this paper we describe newly launched software for on-line Bayesian calibration of archaeological radiocarbon determinations. The software is known as BCal and we invite members of the world-wide archaeological research community to use it should they so wish. All that is required to gain access to the software is a computer connected to the Internet with a modern World-wide Web browser (of the sort you are probably using to read this. BCal does not require access to any additional 'Plug-ins' on your machine. Since the computations needed to obtain the calibrations are undertaken on the BCal server, if you have enough computer power to run your World-wide Web browser you have enough power to use BCal.

  20. Parity Codes Used for On-Line Testing in FPGA

    P. Kubalík


    Full Text Available This paper deals with on-line error detection in digital circuits implemented in FPGAs. Error detection codes have been used to ensure the self-checking property. The adopted fault model is discussed. A fault in a given combinational circuit must be detected and signalized at the time of its appearance and before further distribution of errors. Hence safe operation of the designed system is guaranteed. The check bits generator and the checker were added to the original combinational circuit to detect an error during normal circuit operation. This concurrent error detection ensures the Totally Self-Checking property. Combinational circuit benchmarks have been used in this work in order to compute the quality of the proposed codes. The description of the benchmarks is based on equations and tables. All of our experimental results are obtained by XILINX FPGA implementation EDA tools. A possible TSC structure consisting of several TSC blocks is presented. 

  1. On-line alkali detector based on surface ionization

    Olsson, J.G.; Loenn, B.; Jaeglid, B.; Engvall, K.; Pettersson, J.B.C. [Chalmers University of Technology, Goeteborg (Sweden). Dept. of Physical Chemistry GU


    This project adapts a new on-line alkali measurement technique to coal and biomass combustion and gasification. Alkali metal atoms are known to easily ionize in contact with hot metal surfaces, and the instrument is based on this principle called surface ionization (SI). The primary parts of the detector are a platinum filament and an ion collector. The platinum filament is supported between two electrodes and heated to the temperature for alkali vaporization in ionic form. The ion collector is situated close to the filament. The measured current is proportional to the arrival rate of alkali atoms onto the filament. Laboratory tests were performed on detector sensitivity, detection limit, and time response. Similar sensitivity to both sodium and potassium regardless of molecular form was found. The time response of the detector is determined to be approximately 1 ms enabling it to monitor fast concentration changes in flue gas. Particles with a size below 5 nm melt completely on the hot platinum surface and give similar signals. For larger particles, the ionization efficiency is not 100% and depends on the type of salt. This problem can be overcome with an alternative filament configuration. The detector function was tested in a laboratory high pressure furnace using different fuel samples, atmospheres, and pressures. Alkali release from coal in general is lower than for biomass samples, rate constants and activation energies for alkali release were determined. Measurements were carried out in a biomass pyrolysis apparatus and a gasification pilot plant. The detector function was not influenced by a high concentration of hydrocarbons in the gas phase, and the measurements confirmed detector function in a hostile environment. The detector performed well in laboratory tests, and is a strong candidate for further development into a standard on-line monitor of alkali species in hot flue gas. 10 refs., 16 figs.

  2. Review of Worcestershire On-line Fabric Type Series website

    Beverley Nenk


    Full Text Available The study of archaeological ceramics is advanced through the creation and development of regional and national pottery type-series, which contain samples of each type of pottery identified from a particular area or region. Pottery researchers working in any period, from prehistoric to post-medieval, require access to such type-series, and to their associated data, in order to be able to advance the identification of all types of pottery, not only those types produced in the local area, but those produced in surrounding regions, as well as those imported from abroad. The publication of such type-series, as well as their accessibility to researchers, is essential if the information they contain is to be disseminated. The development of the Worcestershire On-Line Fabric Type Series is the first stage in a remarkable project designed to make the complete fabric and form type series for Worcestershire ceramics accessible on the internet. As part of the Historic Environment Record for Worcestershire, formerly the Sites and Monuments Record, it is designed to improve access to finds and environmental data, with the aim of encouraging and facilitating research. Funded by Worcestershire County Council as part of its commitment to e-government, it is being developed by Worcestershire County Council Archaeology Service with OxfordArchDigital. It is one of a proposed series of on-line specialist resources (to include, for example, clay pipes, environmental archaeology, flint tools, historic buildings, which are also designed to stand alone as research tools. The ceramics website is the first part of Pottery in Perspective, a web-based project to provide information on the pottery used and made in Worcestershire from prehistory to c. 1900AD.

  3. Analysis of a bending test on a full-scale PWR hot leg elbow containing a surface crack

    Delliou, P. le [Electricite de France, EDF, 77 - Moret-sur-Loing (France). Dept. MTC; Julisch, P.; Hippelein, K. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Bezdikian, G. [Electricite de France, EDF, 92 - Paris la Defense (France). Direction Production Transport


    EDF, in co-operation with Framatome, has conducted a large research programme on the mechanical behaviour of thermally aged cast duplex stainless steel elbows, which are part of the main primary circuit of French PWR. One important task of this programme consisted of testing a full-scale PWR hot leg elbow. The elbow contained a semi-elliptical circumferential notch machined on the outer surface of the intrados as well as casting defects located on the flanks. To simulate the end-of-life condition of the component regarding material toughness, it had undergone a 2400 hours ageing heat treatment at 400 C. The test preparation and execution, as well as the material characterization programme, were committed to MPA. The test was conducted under constant internal pressure and in-plane bending (opening mode) at 200 C. For safety reasons, it took place on an open air-site: the Meppen military test ground. At the maximum applied moment (6000 kN.m), the notch did not initiate. This paper presents the experimental results and the fracture mechanics analysis of the test, based on finite element calculations. (orig.)

  4. Integrated functional modeling method for NPP plant DiD risk monitor and its application for conventional PWR

    Yoshikawa, Hidekazu; Yang, Ming; Zhang, Zhijian [Harbin Engineering University, Harbin (China)


    The development of a new risk monitor system is introduced in this paper, which can be applied not only to severe accident prevention in daily operation but also to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. The summary of the fundamental method is summarized on how to configure the Plant Defense in-Depth (Did) Risk Monitor by object-oriented software system based on functional modeling approach. Following the authors??preceding preliminary study for AP1000, the way of realizing the proposed method of configuring the plant Did risk monitor was investigated for a safety-enhanced Japanese PWR design to meet with the tight anti-severe accident requirements set by national regulation in Japan after Fukushima Daiichi accident. The result of this example practice of the presented preliminary study for Japanese PWR was for the level 4 of the Did in case of beyond design basis accident, that is, loss of all AC power + RCP seal LOCA, against the former case of AP1000 for level 3 Did in case of large LOCA.

  5. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)


    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

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

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


    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.

  7. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    Mayers, J.B.; Soth, L.G.


    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.

  8. Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

    M. Avramova


    Full Text Available The Pennsylvania State University (PSU under the sponsorship of the US Nuclear Regulatory Commission (NRC has prepared, organized, conducted, and summarized the Organisation for Economic Co-operation and Development/US Nuclear Regulatory Commission (OECD/NRC benchmark based on the Nuclear Power Engineering Corporation (NUPEC pressurized water reactor (PWR subchannel and bundle tests (PSBTs. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency (NEA of OECD and the Japan Nuclear Energy Safety Organization (JNES, Japan. The OECD/NRC PSBT benchmark was organized to provide a test bed for assessing the capabilities of various thermal-hydraulic subchannel, system, and computational fluid dynamics (CFDs codes. The benchmark was designed to systematically assess and compare the participants’ numerical models for prediction of detailed subchannel void distribution and department from nucleate boiling (DNB, under steady-state and transient conditions, to full-scale experimental data. This paper provides an overview of the objectives of the benchmark along with a definition of the benchmark phases and exercises. The NUPEC PWR PSBT facility and the specific methods used in the void distribution measurements are discussed followed by a summary of comparative analyses of submitted final results for the exercises of the two benchmark phases.

  9. The Effects of Hot Bending on the Low Cycle Fatigue Behaviors of 347 SS in PWR Primary Environment

    Kim, Ho-Sub; Hong, Jong-Dae; Lee, Junho; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)


    Fatigue damage could be significant for some locations, especially the welds and bends where stress concentration is typically high. As a possible solution, a large radius hot-bending method has been suggested to eliminate some weld joints and all tight bends. However, for the hot-bending process which involves a high temperature thermal cycle, there is a concern about changes in mechanical properties including low cycle fatigue behaviors. In APR1400, Type 347 SS have been used as surge line pipes. Therefore, to verify the applicability of hot-bending on 347 SS surge line pipes, an environmental fatigue test program was initiated. In this paper, the preliminary results of the on-going test program are introduced. Also, the low cycle fatigue behaviors of 347 SS are compared with those of other grade of stainless steels. The effects of hot bending on the low cycle fatigue behavior of 347 SS were quantitatively evaluated. The fatigue life was compared with the estimated values per NUREG 6909 rev. 1. There are no distinct differences between NUREG 6909 and LCF tests. According to fractography and cross section analysis in progress, basically, the reduction of LCF life of 347 SS in PWR water was caused by operation of HIC mechanism. The cyclic stress responses shows that there is no secondary hardening in 330 .deg.C air and PWR water.

  10. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Andrej Prošek


    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  11. Determination of anions with an on-line capillary electrophoresis method; Anionien on-line maeaeritys kapillaarielektroforeesilla - MPKT 10

    Siren, H.; Saerme, T.; Kotiaho, T.; Hiissa, T.; Savolahti, P.; Komppa, V. [VTT Chemical Technology, Espoo (Finland)


    The aim of the study was to set-up an on-line capillary electrophoresis method for determination of anions in process waters of pulp and paper industry with exporting the results to the process control system of the mill. The quantification is important, since it will give information about the possible causes of precipitation. In recent years, the capillary electrophoresis (CE) due to its high separation efficiency has been shown as a method to take into consideration when analyzing chemical species ranging from small inorganic anions to different macromolecules. Many compounds are not easily detected in their native state, why analysis methods must be developed to improve their detection. Especially, small inorganic and organic anions which do not have chromophores are not sensitive enough for direct-UV detection. In such analyses the anions are mostly detected with indirect-UV technique. Capillary electrophoresis instruments are used to analyze samples in off-line, which seldom represent the situation in process. Therefore, on-line instrument technology with autoanalyzing settings will be needed in quality control. The development of a fully automatic capillary electrophoresis system is underway in co-operation with KCL (The Finnish Pulp and Paper Research Institute). In our research, we have first concentrated on the determination of sulphate in waters of paper industry. The method used for detection of sulphate is based on indirect-UV detection with CE, where the background electrolyte (BGE) is an absorbing mixture of secondary amines. The whole procedure for quantification of sulphate is performed within 15 minutes, after which a new sample is analyzed automatically. The only sample pretreatment is filtration, which is necessary before analysis. The concentrations of sulphate in process waters tested were between 300 and 800 ppm. Our tests show that a simultaneous determination of chloride, sulphate, nitrate, nitrite, sulphite, carbonate and oxalate is also

  12. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen; Zhou, Bangxin; Shoji, Tetsuo


    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T-L and L-T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T-L orientation with a higher crack growth rate than that in the specimen L-T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L-T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant.

  13. Example Calculations of In{sub v}essel Steam Explosions for a Prototypical PWR

    Park, Ik Kyu; Hong, Seong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    In this paper, the sample calculation for the in{sub v}essel steam explosions were done by using the MC3D code. The evaluation of the computational code had been done against TROI experiments and the code had been adapted to a PWR ex{sub v}essel steam explosion calculations. MC3D is a code for the calculation of different types of multiphase multi-component flows. It has been built with the fuel-coolant interaction calculations in mind. It is, however, able to calculate very different situations and has a rather wide field of potential applications. MC3D is a set of two fuel-coolant interaction codes with a common numeric solver, one for the premixing phase and one for the explosion phase. In general, the steam explosion simulation with MC3D is being carried out in two steps. In the first step, the distributions of the melt, water, and vapor phases at steam explosion triggering are being calculated with the premixing module. These premixing simulation results present the input for the second step when the escalation and propagation of the steam explosion through the premixture are being calculated with the explosion module. The MC3D premixing model is a six-field application in which the melt is described by three fields. The first one is called 'continuous' and can describe many situations as, e.g., a jet or the melt lying on the bottom of a vessel. The second field corresponds to the droplets issued from the jet fragmentation. This field describes the discontinuous state of the fuel. The third field is optional and describes the fuel fragments issuing from drop fine fragmentation. The remaining three fields are the water, the vapor, and a noncondensable gas. The drop surface area is modeled with a standard interfacial area transport equation. In the explosion model, the continuous phase is not present and only the two fields related to the dispersed fuel are considered

  14. Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

    Yang, Zijiang; Shan, Jianqiang, E-mail:; Gou, Junli


    Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

  15. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    None, None


    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  16. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    None, None


    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  17. 77 FR 55811 - Manufacturing Extension Partnership Advisory Board On-line Open Meeting


    ... National Institute of Standards and Technology Manufacturing Extension Partnership Advisory Board On-line... Manufacturing Extension Partnership (MEP) Advisory Board will hold an open, on-line meeting via webcast on... their remote location. Questions regarding the on-line meeting should be sent to the...

  18. Women Physicians Are Early Adopters of On-Line Continuing Medical Education

    Harris, John M., Jr.; Novalis-Marine, Cheryl; Harris, Robin B.


    Introduction: On-line continuing medical education (CME) provides advantages to physicians and to medical educators. Although practicing physicians increasingly use on-line CME to meet their educational needs, the overall use of on-line CME remains limited. There are few data to describe the physicians who use this new educational medium; yet,…

  19. Development of on-line laser power monitoring system

    Ding, Chien-Fang; Lee, Meng-Shiou; Li, Kuan-Ming


    Since the laser was invented, laser has been applied in many fields such as material processing, communication, measurement, biomedical engineering, defense industries and etc. Laser power is an important parameter in laser material processing, i.e. laser cutting, and laser drilling. However, the laser power is easily affected by the environment temperature, we tend to monitor the laser power status, ensuring there is an effective material processing. Besides, the response time of current laser power meters is too long, they cannot measure laser power accurately in a short time. To be more precisely, we can know the status of laser power and help us to achieve an effective material processing at the same time. To monitor the laser power, this study utilize a CMOS (Complementary metal-oxide-semiconductor) camera to develop an on-line laser power monitoring system. The CMOS camera captures images of incident laser beam after it is split and attenuated by beam splitter and neutral density filter. By comparing the average brightness of the beam spots and measurement results from laser power meter, laser power can be estimated. Under continuous measuring mode, the average measuring error is about 3%, and the response time is at least 3.6 second shorter than thermopile power meters; under trigger measuring mode which enables the CMOS camera to synchronize with intermittent laser output, the average measuring error is less than 3%, and the shortest response time is 20 millisecond.

  20. AMADEUS on-line trigger and filtering methods

    Neff, M. [Erlangen Centre for Astroparticle Physics (ECAP), Friedrich-Alexander-Universitaet Erlangen-Nuernberg, Physikalisches Institut, Erwin-Rommel-Strasse 1, D-91058 Erlangen (Germany)], E-mail:; Anton, G.; Graf, K.; Hoessl, J.; Katz, U.; Lahmann, R.; Richardt, C. [Erlangen Centre for Astroparticle Physics (ECAP), Friedrich-Alexander-Universitaet Erlangen-Nuernberg, Physikalisches Institut, Erwin-Rommel-Strasse 1, D-91058 Erlangen (Germany)


    AMADEUS is a system designed to investigate the method of acoustic particle detection for high energy neutrinos and the acoustic environment in the deep sea as part of the ANTARES neutrino telescope. In this context, six local clusters of six acoustic sensors each were integrated into the ANTARES infrastructure. The first three clusters have been taking data since December 2007 and the second three since the completion of ANTARES in May 2008. In the paper, the methods used for the on-line triggering and filtering of the data acquired with the AMADEUS set-up are described. On-shore, a dedicated computer-cluster is used to control the off-shore DAQ hardware, to process and store the acoustic data arriving from the sensors. On this cluster different data filtering schemes and triggers are implemented. Transient signals are selected by a variable threshold, which is self-adjusting to the changing conditions of the deep sea. For bipolar pulses-the characteristic acoustic signature of a neutrino-a pattern recognition is used based on cross-correlating the output of the sensors with a pre-defined bipolar pulse. To study the characteristics of the ambient noise in the deep sea an amount of unfiltered data is stored in regular intervals.

  1. On-Line Condition Monitoring using Computational Intelligence

    Vilakazi, C B; Mautla, P; Moloto, E


    This paper presents bushing condition monitoring frameworks that use multi-layer perceptrons (MLP), radial basis functions (RBF) and support vector machines (SVM) classifiers. The first level of the framework determines if the bushing is faulty or not while the second level determines the type of fault. The diagnostic gases in the bushings are analyzed using the dissolve gas analysis. MLP gives superior performance in terms of accuracy and training time than SVM and RBF. In addition, an on-line bushing condition monitoring approach, which is able to adapt to newly acquired data are introduced. This approach is able to accommodate new classes that are introduced by incoming data and is implemented using an incremental learning algorithm that uses MLP. The testing results improved from 67.5% to 95.8% as new data were introduced and the testing results improved from 60% to 95.3% as new conditions were introduced. On average the confidence value of the framework on its decision was 0.92.


    Marlise Horst


    Full Text Available University students used a set of existing and purpose-built on-line tools for vocabulary learning in an experimental ESL course. The resources included concordance, dictionary, cloze-builder, hypertext, and a database with interactive self-quizzing feature (all freely available at The vocabulary targeted for learning consisted of (a Coxhead's (2000 Academic Word List, a list of items that occur frequently in university textbooks, and (b unfamiliar words students had met in academic texts and selected for entry into the class database. The suite of tools were designed to foster retention by engaging learners in deep processing, an aspect that is often described as missing in computer exercises for vocabulary learning. Database entries were examined to determine whether context sentences supported word meanings adequately and whether entered words reflected the unavailability of cognates in the various first languages of the participants. Pre- and post-treatment performance on tests of knowledge of words targeted for learning in the course were compared to establish learning gains. Regression analyses investigated connections between use of specific computer tools and gains.

  3. On-line range images registration with GPGPU

    Będkowski, J.; Naruniec, J.


    This paper concerns implementation of algorithms in the two important aspects of modern 3D data processing: data registration and segmentation. Solution proposed for the first topic is based on the 3D space decomposition, while the latter on image processing and local neighbourhood search. Data processing is implemented by using NVIDIA compute unified device architecture (NIVIDIA CUDA) parallel computation. The result of the segmentation is a coloured map where different colours correspond to different objects, such as walls, floor and stairs. The research is related to the problem of collecting 3D data with a RGB-D camera mounted on a rotated head, to be used in mobile robot applications. Performance of the data registration algorithm is aimed for on-line processing. The iterative closest point (ICP) approach is chosen as a registration method. Computations are based on the parallel fast nearest neighbour search. This procedure decomposes 3D space into cubic buckets and, therefore, the time of the matching is deterministic. First technique of the data segmentation uses accele-rometers integrated with a RGB-D sensor to obtain rotation compensation and image processing method for defining pre-requisites of the known categories. The second technique uses the adapted nearest neighbour search procedure for obtaining normal vectors for each range point.

  4. The on-line low temperature nuclear orientation facility NICOLE

    Ohtsubo, T.; Roccia, S.; Stone, N. J.; Stone, J. R.; Gaulard, C.; Köster, U.; Nikolov, J.; Simpson, G. S.; Veskovic, M.


    We review major experiments and results obtained by the on-line low temperature nuclear orientation method at the NICOLE facility at ISOLDE, CERN since the year 2000 and highlight their general physical impact. This versatile facility, providing a large degree of controlled nuclear polarization, was used for a long-standing study of magnetic moments at shell closures in the region Z = 28, N = 28–50 but also for dedicated studies in the deformed region around A ∼ 180. Another physics program was conducted to test symmetry in the weak sector and constrain weak coupling beyond V–A. Those two programs were supported by careful measurements of the involved solid state physics parameters to attain the full sensitivity of the technique and provide interesting interdisciplinary results. Future plans for this facility include the challenging idea of measuring the beta–gamma–neutron angular distributions from polarized beta delayed neutron emitters, further test of fundamental symmetries and obtaining nuclear structure data used in medical applications. The facility will also continue to contribute to both the nuclear structure and fundamental symmetry test programs.

  5. The Efficiency of the On-line Samplings

    Ileana Gabriela NICULESCU-ARON


    Full Text Available The rapid growth of the technology from the last decades led to the collateral development of many other sciences. One of the most important inventions was the Internet and the web technologies with a tremendous impact on the society. Statistics, as a social science, at its turn in ongoing development has only to gain from that. Lately, the on line sampling technique greatly developed. Each web site of a certain importance includes in various forms of the questionnaires. These vary from a mere question to lengthy ones and are a part of daily life of those who access the World Wide Web. The main question is how feasible are the results derived from these samplings as the main issue is the representativiness. A nonrepresentative sampling is a futile one. It is a more convenient solution to post a question on the web page and to wait for an answer from the page’s visitors? But how representative is this answer for the target audience? The present paper aims to list the on-lone methodology as well as analyze their efficiency through presenting their advantages and drawbacks.

  6. Using on-line altered auditory feedback treating Parkinsonian speech

    Wang, Emily; Verhagen, Leo; de Vries, Meinou H.


    Patients with advanced Parkinson's disease tend to have dysarthric speech that is hesitant, accelerated, and repetitive, and that is often resistant to behavior speech therapy. In this pilot study, the speech disturbances were treated using on-line altered feedbacks (AF) provided by SpeechEasy (SE), an in-the-ear device registered with the FDA for use in humans to treat chronic stuttering. Eight PD patients participated in the study. All had moderate to severe speech disturbances. In addition, two patients had moderate recurring stuttering at the onset of PD after long remission since adolescence, two had bilateral STN DBS, and two bilateral pallidal DBS. An effective combination of delayed auditory feedback and frequency-altered feedback was selected for each subject and provided via SE worn in one ear. All subjects produced speech samples (structured-monologue and reading) under three conditions: baseline, with SE without, and with feedbacks. The speech samples were randomly presented and rated for speech intelligibility goodness using UPDRS-III item 18 and the speaking rate. The results indicted that SpeechEasy is well tolerated and AF can improve speech intelligibility in spontaneous speech. Further investigational use of this device for treating speech disorders in PD is warranted [Work partially supported by Janus Dev. Group, Inc.].

  7. Chromatographic on-line detection of bioactives in food

    Remmelt Van der Werf


    Full Text Available ABSTRACTFindings were focused on the anti-oxidative activity of numerous fruits and vegetables by means of an on-line HPLC radical scavenging detection method. The reactant used was the ABTS•+ green radical cation. The system has been optimized in terms of reactor design, and chemical reactions kinetics. It has been qualified to classify molecules in order of their increasing activity to scavenge exogenous radicals. It may be used as a powerful high resolution screening tool to investigate the radical scavenging activities of natural plants. Bioassays consisting in cellular in vitro antioxidant assay using pancreatic β-cells have been used to confirm the bioactivity of the selected micronutrients. This study demonstrated that it is possible to screen at the molecular level, the bioactivity of numerous natural samples and to point out the richness of the local biodiversity in terms of natural resource of functional food ingredients usable for their potential benefits for consumer’s health, wellbeing and wellaging.Key words: HPLC radical scavenging detection method, bioactivity of natural samples

  8. On-line catalytic upgrading of biomass fast pyrolysis products

    LU Qiang; ZHU XiFeng; LI WenZhi; ZHANG Ying; CHEN DengYu


    Pyrolysis-gas chromatography/mass spectrometry (Py-GC/MS) was employed to achieve fast pyrolysis of biomass and on-line analysis of the pyrolysis vapors. Four biomass materials (poplar wood, fir wood, cotton straw and rice husk) were pyrolyzed to reveal the difference among their products. Moreover, catalytic cracking of the pyrolysis vapors from cotton straw was performed by using five catalysts, including two microporous zeolites (HZSM-5 and HY) and three mesoporous catalysts (ZrO2&TiO2, SBA-15 and AI/SBA-15). The results showed that the distribution of the pyrolytic products from the four materials differed a little from each other, while catalytic cracking could significantly alter the pyrolytic products. Those important primary pyrolytic products such as levoglucosen, hydroxyacetaldehyde and 1-hydroxy-2-propanone were decreased greatly after catalysis. The two microporous zeolites were ef-fective to generate high yields of hydrocarbons, while the three mesoporous materials favored the formation of furan, furfural and other furan compounds, as well as acetic acid.

  9. Virtual laboratories: Collaborative environments and facilities-on-line

    Thomas, C.E. Jr. [Oak Ridge National Lab., TN (United States). I and C Div.; Cavallini, J.S.; Seweryniak, G.R.; Kitchens, T.A.; Hitchcock, D.A.; Scott, M.A.; Welch, L.C. [Dept. of Energy, Germantown, MD (United States). Mathematical Information, and Computational Sciences Div.; Aiken, R.J. [Dept. of Energy, Germantown, MD (United States). Mathematical Information, and Computational Sciences Div.]|[Lawrence Livermore National Lab., CA (United States); Stevens, R.L. [Argonne National Lab., IL (United States). Mathematics and Computer Sciences Div.


    The Department of Energy (DOE) has major research laboratories in a number of locations in the US, typically co-located with large research instruments or research facilities valued at tens of millions to even billions of dollars. Present budget exigencies facing the entire nation are felt very deeply at DOE, just as elsewhere. Advances over the last few years in networking and computing technologies make virtual collaborative environments and conduct of experiments over the internetwork structure a possibility. The authors believe that development of these collaborative environments and facilities-on-line could lead to a ``virtual laboratory`` with tremendous potential for decreasing the costs of research and increasing the productivity of their capital investment in research facilities. The majority of these cost savings would be due to increased productivity of their research efforts, better utilization of resources and facilities, and avoiding duplication of expensive facilities. A vision of how this might all fit together and a discussion of the infrastructure necessary to enable these developments is presented.

  10. Gas-liquid countercurrent two-phase flow in a PWR hot leg: A comprehensive research review

    Deendarlianto, E-mail: [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Department of Mechanical and Industrial Engineering, Faculty of Engineering, Gadjah Mada University, Jalan Grafika No. 2, Yogyakarta 55281 (Indonesia); Hoehne, Thomas; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Vierow, Karen [Department of Nuclear Engineering Texas A and M University, 129 Zachry Engineering Center, 3133 TAMU College Station, TX 77843-3133 (United States)


    Highlights: Black-Right-Pointing-Pointer We review the scientific progress on the CCFL in a PWR hot leg. Black-Right-Pointing-Pointer It includes the experimental data, one-dimensional and CFD models in the open literatures. Black-Right-Pointing-Pointer The weak and strong points of the published works were clarified. Black-Right-Pointing-Pointer The research directions in this field were proposed. - Abstract: Research into gas-liquid countercurrent two-phase flow in a model of pressurized water reactor (PWR) hot leg has been carried out over the last several decades. An extensive experimental data base has been accumulated from these studies, leading to the development of phenomenological correlations and scaling parameters of the countercurrent flow limitation (CCFL). However, most of the proposed correlations apply under a relatively narrow range of conditions, generally limited to the test section conditions and/or geometry. Moreover the development of mechanistic models based on the underlying physical processes has been limited. In contrast to this mechanistic form of modelling, the implementation of computational fluid dynamics (CFD) techniques has also been pursued, but the considerable robust three-dimensional (3D) closure relations for this application remain an unachieved goal due to lack of detailed phenomenological knowledge and consequent application of empirical one-dimensional experimental correlations to the multidimensional problem. This paper presents a comprehensive review of research work on countercurrent gas-liquid two-phase flow in a PWR hot leg and provides direction regarding future research on this topic. In the introductory section, the problems facing current research are described. In the following sections, recent experimental as well as theoretical research achievements are overviewed. In the last section, the problems that remain unsolved are discussed, along with some concluding remarks. It was found that only limited theoretical

  11. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail:


    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  12. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  13. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method; Sondas para inspecao de trocadores de calor instalados em usinas nucleares tipo PWR pelo metodo de correntes parasitas

    Silva, Alonso F.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Enghenharia Mecanica]. E-mail:; Alencar, Donizete A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail:


    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

  14. Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR; Evaluacion de la presencia de un absorbedor quemable en un ensamble 3x3 tipo PWR

    Martinez F, M. A.; Del Valle G, E.; Alonso V, G. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico)]. e-mail:


    In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO{sub 2}. (Author)

  15. Worldwide On-line Distance Learning University Astronomy

    Eyres, S. P. S.; Hassall, B. J. M.; Butchart, I.; Bromage, G. E.


    The University of Central Lancashire operates a suite of distance learning courses in Astronomy, available both on-line and via CD-ROM. The courses are available worldwide, and emphasise flexibility of study. To this end students can study anything from a single module (1/6^th of a full year at degree level) all the way up to an entire degree entirely by distance learning. Study rates vary from one to four modules each year, and students can move on to Level 2 modules (equivalent to second year level in a UK degree) before completing the full set of Level 1 modules. Over 1000 awards have been made to date. The core syllabus is Astronomy and Cosmology at Level 1, alongside skills in literature research, using computers, and basic observing. We also offer a basic history of European astronomy. At Level 2 we look at the astrophysics of the Sun, the stars, and galaxies including the Milky Way. By Level 3 students are expected to engage in a large individual project, and a collaborative investigation with other students, alongside high-level courses in cosmology, relativity, extreme states of matter and the origins of the elements, life and astronomical objects. While many students are retired people looking to exercise their brains, keen amateur amateurs or professionals with disposable incomes, and significant fraction are teachers seeking to improve their subject knowledge or high school students gaining an edge in the UK University entrance competition. Via our involvement with SALT we offer our courses to members of previously disadvantaged communities. This leads to an incredibly diverse and lively student body.

  16. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)


    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  17. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Hartini, Entin, E-mail:; Andiwijayakusuma, Dinan, E-mail: [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)


    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  18. The development and verification of thermal-hydraulic code on passive residual heat removal system of Chinese advanced PWR


    The technology of passive safety is the current trend among safety systems in nuclear power plant. Passive residual heat removal system (PRHRS), a major part of passive safety systems of Chinese advanced PWR, is a novel design with three-fold natural circulation. On the basis of reasonable physics and mathematics models, MITAP-PRHRS code was developed to analyze steady and transient characteristics of the PRHRS. The calculation and analysis show that the code simulates steady characteristics of the PRHRS very well, and it is able to simulate transient characteristics of all startup modes of the PRHRS. However, the quantitative description is poor during the initial stages of the transition process when water hammer occurs.

  19. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan


    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  20. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail:


    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  1. Valve inlet fluid conditions for pressurizer safety and relief valves in combustion engineering-designed plants. Final report. [PWR

    Bahr, J.; Chari, D.; Puchir, M.; Weismantel, S.


    The purpose of this study is to assemble documented information for C-E designed plants concerning pressurizer safety and power operated relief valve (PROV) inlet fluid conditions during actuation as calculated by conventional licensing analyses. This information is to be used to assist in the justification of the valve inlet fluid conditions selected for the testing of safety valves and PORVs in the EPRI/PWR Safety/Relief Valve Test Program. Available FSAR/Reload analyses and certain low temperature overpressurization analyses were reviewed to identify the pressurization transients which would actuate the valves, and the corresponding valve inlet fluid conditions. In addition, consideration was given to the Extended High Pressure Liquid Injection event. A general description of each pressurization transient is provided. The specific fluid conditions identified and tabulated for each C-E designed plant for each transient are peak pressurizer pressure, pressure ramp rate at actuation, temperature and fluid state.

  2. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    Hartini, Entin; Andiwijayakusuma, Dinan


    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  3. PWR neutron ex-vessel detection calculations using three-dimensional codes; Calculs de detection neutronique externe dans un rep

    Dekens, O.; Lefebvre, J.C.; Rohart, M. [Electricite de France (EDF), 69 -Villeurbanne (France); Chiron, M. [CEA Centre d`Etudes de Saclay, 91 -Gif-sur-Yvette (France). Direction des Reacteurs Nucleaires; Wouters, R. de [TRACTEBEL, Brussels (Belgium)


    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l`Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors). 11 refs.

  4. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    McCardell, R.K.; MacDonald, P.E.


    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  5. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    Kelly, G N


    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  6. Practical Application of the MFM Suite on a PWR System: Modelling and Reasoning on Causes and Consequences of Process Anomalies

    Zhang, Xinxin; Thunem, Harald P - J; Lind, Morten


    Multilevel Flow Modelling (MFM) is a functional modelling methodology which applies means - end and parts - whole decomposition and aggregation techniques to handle the complexity of engineering systems. It has been adopted in several case studies to model the process goal and functions of PWR...... is equipped with an MFM Model Editing Interface to facilitate the modelling process and MFM model analysis modules to run diag nosis and prognosis analyses based on developed models. New features of the MFM Suite also include making corresponding process diagram for the plant being modelled with MFM...... and linking the MFM model to its process components. The purpose of this report is to make a comprehensive demonstration of how to use the MFM Suite to develop MFM models and run causal reasoning for abnormal situations. This report will explain the capability of representing process and operational knowledge...

  7. ASTEC V2.0 reactor applications on French PWR 900 MWe accident sequences and comparison with MAAP4

    Lombard, Virginie; Azarian, Garo; Ducousso, Erik; Gandrille, Pascal, E-mail:


    In the frame of the SARNET Severe Accident Network of Excellence an important task of partners is the assessment of the ASTEC integral code, considered today as the European reference code for evaluation of the source term. A code-to-code comparison between ASTEC V2.0 rev1 and MAAP 4.0.7 code versions has been performed by AREVA NP SAS on a French PWR 900 MWe. Two transients have been analyzed, focussing on in-vessel phenomena: total loss of feedwater (H2 sequence in the French nomenclature) and total loss of onsite and offsite power (H3 sequence). The detailed analysis shows an overall good agreement between both code results on thermal-hydraulics, hydrogen production and core degradation phenomena.


    Sri Kuntjoro


    Full Text Available Penambahan pembangkit listrik yang baru khususnya pembangkit listrik tenaga nuklir (PLTN berpotensi memberikan konsekuensi radiologis pada masyarakat dan lingkungan, karena adanya lepasan radioaktif dalam kondisi operasi normal maupun abnormal. Oleh karena itu maka pengelola reaktor nuklir harus bisa menyediakan data dan argumentasi yang kuat untuk menjelaskan tentang keselamatan PLTN terhadap lingkungan. Untuk itu perlu dilakukan analisis kondisi abnormal yang terjadi pada PLTN yang akan memberikan konsekuensi radiologis pada lingkungan. Analisis dilakukan dengan membuat pemodelan simulasi kondisi abnormal yang dipostulasikan pada PLTN tipe PWR 1000 MWe serta simulasi dan pemodelan pola potensi lingkungan sebagai daya dukung tapak terhadap penerimaan konsekuensi radiologis tersebut. Pemodelan fenomena transport radionuklida dari teras reaktor sampai ke luar dari sungkup reaktor dilakukan menggunakan perangkat lunak EMERALD dan pemodelan pola dispersi radioaktivitas ke lingkungan dari reaktor meliputi simulasi kondisi meteorologi, distribusi penduduk, produksi dan konsumsi masyarakat pada kondisi ekstrim di daerah studi, menggunakan perangkat lunak GIS, Arcview, Windrose, dan PC COSYMA. Pemodelan konsekuensi radiologis menggunakan tapak contoh daerah Bojanegara-Kramatwatu Pantai Serang-Banten. Dengan menggunakan data sourceterm, data meteorologi dan data dispersi (sebaran penduduk, produksi pertanian dan ternak dan modeling alur paparan (pathway, dihasilkan model sebaran radionuklida dan penerimaan paparan radiasi di lingkungan tapak Bojanegara-Serang, dengan penerimaan dosis radiasi di bawah batas yang diijinkan badan regulator BAPETEN. Kata kunci : PLTN, radioaktivitas, pola dispersi, keselamatan   Additional of electrical power especially Nuclear Power Plant will give radiological consequences to population and environment due to radioactive release in normal and abnormal condition. In consequence the management of nuclear power plant must

  9. Multi level optimization of burnable poison utilization for advanced PWR fuel management

    Yilmaz, Serkan

    The objective of this study was to develop an unique methodology and a practical tool for designing burnable poison (BP) pattern for a given PWR core. Two techniques were studied in developing this tool. First, the deterministic technique called Modified Power Shape Forced Diffusion (MPSFD) method followed by a fine tuning algorithm, based on some heuristic rules, was developed to achieve this goal. Second, an efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poisons (BPs) placement optimization problem for a reference Three Mile Island-1 (TMI-1) core. This thesis presents the step by step progress in developing such a tool. The developed deterministic method appeared to perform as expected. The GA technique produced excellent BP designs. It was discovered that the Beginning of Cycle (BOC) Kinf of a BP fuel assembly (FA) design is a good filter to eliminate invalid BP designs created during the optimization process. By eliminating all BP designs having BOC Kinf above a set limit, the computational time was greatly reduced since the evaluation process with reactor physics calculations for an invalid solution is canceled. Moreover, the GA was applied to develop the BP loading pattern to minimize the total Gadolinium (Gd) amount in the core together with the residual binding at End-of-Cycle (EOC) and to keep the maximum peak pin power during core depletion and Soluble boron concentration at BOC both less than their limit values. The number of UO2/Gd2O3 pins and Gd 2O3 concentrations for each fresh fuel location in the core are the decision variables and the total amount of the Gd in the core and maximum peak pin power during core depletion are in the fitness functions. The use of different fitness function definition and forcing the solution movement towards to desired region in the solution space accelerated the GA runs. Special emphasize is given to minimizing the residual binding to increase core lifetime as

  10. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Wu, Yichun, E-mail: [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: [College of Energy, Xiamen University, Xiamen 361102 (China)


    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  11. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    LaFleur, Adrienne M., E-mail:; Menlove, Howard O., E-mail:


    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  12. Effects of cold work and stress on oxidation and SCC behavior of stainless steels in PWR primary water environments

    Shoji, T.; Sakaguchi, K.; Lu, Z. [Fracture and Reliability Research Institute, Tohoku University, Sendai City 980-8579 (Japan); Hirano, S.; Hasegawa, Y. [Kansai Electric Power Co (Japan); Kobayashi, T.; Fujimoto, K.; Nomura, Y. [Mitsubishi Heavy Industries (Japan)


    Intergranular stress corrosion cracking (SCC) samples taken from a weld HAZ of 316 stainless steel welded to a low alloy steel of steam generator nozzle with nickel base alloy 82 in Mihama Unit 2 PWR plant were analyzed by extensive metallographic observation, micro-Raman spectroscopy, TEM analysis of stainless steel material, oxide morphology, compositional profiles as well as their crystal structures. The crack growth history during the plant operation is discussed in connection to a residual stress distribution at HAZ and distribution of oxides on/in the cracks. Possible time dependence of crack growth rate with crack growth in components was proposed based upon the evidences observed about oxides. The importance of surface integrity assessment in SCC initiation and propagation is emphasized from a point of view of oxidation localization which can be promoted by strain (dislocation density), straining and stress, which play a crucial role in oxidation due to accelerated mass transfer in oxides as well as underlying metallic materials. Especially, preferential oxidation along slip bands suggests that oxygen diffusion in such a region with a high dislocation density is faster than the other region. This fact implies that grain boundary can also be a preferential path of oxidation as has been observed by TEM, TOFSIMS and 3D-APT. This localization of oxidation and acceleration is discussed based upon an analysis of profile development at a stressed oxide/metal interface. The effects of environmental parameters, temperature, loading mode, and rolling procedures on SCC of stainless steels in simulated PWR environments were investigated by laboratory tests. Strong interactions among grain boundary structure, environmental parameters and interfacial oxidation kinetics, and SCC behavior are observed

  13. On-line Geoscience Data Resources for Today's Undergraduates

    Goodwillie, A. M.; Ryan, W.; Carbotte, S.; Melkonian, A.; Coplan, J.; Arko, R.; O'Hara, S.; Ferrini, V.; Leung, A.; Bonckzowski, J.


    Broadening the experience of undergraduates can be achieved by enabling free, unrestricted and convenient access to real scientific data. With funding from the U.S. National Science Foundation, the Marine Geoscience Data System (MGDS) ( serves as the integrated data portal for various NSF-funded projects and provides free public access and preservation to a wide variety of marine and terrestrial data including rock, fluid, biology and sediment samples information, underway geophysical data and multibeam bathymetry, water column and multi-channel seismics data. Users can easily view the locations of cruise tracks, sample and station locations against a backdrop of a multi-resolution global digital elevation model. A Search For Data web page rapidly extracts data holdings from the database and can be filtered on data and device type, field program ID, investigator name, geographical and date bounds. The data access experience is boosted by the MGDS use of standardised OGC-compliant Web Services to support uniform programmatic interfaces. GeoMapApp (, a free MGDS data visualization tool, supports map-based dynamic exploration of a broad suite of geosciences data. Built-in land and marine data sets include tectonic plate boundary compilations, DSDP/ODP core logs, earthquake events, seafloor photos, and submersible dive tracks. Seamless links take users to data held by external partner repositories including PetDB, UNAVCO, IRIS and NGDC. Users can generate custom maps and grids and import their own data sets and grids. A set of short, video-style on-line tutorials familiarises users step- by-step with GeoMapApp functionality ( Virtual Ocean ( combines the functionality of GeoMapApp with a 3-D earth browser built using the NASA WorldWind API for a powerful new data resource. MGDS education involvement (, go to Education tab

  14. Semi-on-line analysis for fast and precise monitoring of bioreaction processes

    Christensen, L.H.; Marcher, J.; Schulze, Ulrik


    Monitoring of substrates and products during fermentation processes can be achieved either by on-line, in situ sensors or by semi-on-line analysis consisting of an automatic sampling step followed by an ex situ analysis of the retrieved sample. The potential risk of introducing time delays...... and signal bias during sampling makes it necessary to distinguish between real-time, on-line, in situ methods and semi-on-line analysis. In addition, semi-on-line analyzers are often mechanically complex-a circumstance which has to be given special attention during their industrial use on a routine basis....... This review on semi-on-line analysis will focus both on the dynamics and precision of aseptic sampling devices and on the performance of flow injection analysis (FIA) and sequential injection analysis (SIA), especially with regard to their robustness when used in industry. (C) 1996 John Wiley & Sons, Inc....

  15. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Bocanegra, R.; Jimenez, G.


    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  16. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.


    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  17. On-line determination of anions in pulp mills by capillary electrophoresis (CE); Tehdasoloissa tapahtuva anionien kapillaarielektroforeettinen on-line maeaeritys ja sen hyoedyntaeminen prosessivalvonnassa - MPKY 02

    Kokkonen, R.; Holmberg, M.; Vainikka, V. [Finnish Pulp and Paper Research Institute, Espoo (Finland)


    The aim of the study was to set-up a process control system for on-line measurement of certain anions. Typical anions which forms precipitates in pulp and paper mills are oxalate, carbonate and sulphate. Thus it is important to develop a continuous process analyzing system to control concentration levels of this anions. For the preliminary tests of continuous determinations of chloride and sulphate anions in tap water a simple on-line system was build in KCL (The Finnish Pulp and Paper Research Institute) and connected to a capillary electroforesis apparatus. In the preliminary tests a chromate buffer (ph = 7.6) was used. Separation of chloride and sulphate was excellent but the stability of buffer was not good enough and it was usable only for few hours. After experimental studies VTT developed a stable capillary electrophoresis method based on mixed amine buffer and this was selected for an on-line method for determination of anions in process waters of the pulp and paper industry. In the preliminary on-line test (r = 20) repeatabilities of migration times of sulphate and chloride with the chromate buffer were < 5 % (RSD) and peak heights < 15 % (RSD). With the mixed amine buffer repeatabilities were better. The preliminary tests showed that it is possible to connect a capillary electrophoresis system to on-line measurements. For the moment no commercial on-line CE apparatus is available. (orig.)

  18. On-line Scheduling Algorithm for Penicillin Fed-batch Fermentation

    XUE Yao-feng; YUAN Jing-qi


    An on-line scheduling algorithm to maximize gross profit of penicillin fed-batch fermentation is proposed. According to the on-line classification method, fed-batch fermentation batches are classified into three categories. Using the scheduling strategy, the optimal termination sequence of batches is obtained. Pseudo on-line simulations for testing the proposed algorithm with the data from industrial scale penicillin fermentation are carried out.

  19. Assessment of subcriticality during PWR-type reactor refueling; Evaluation de la sous-criticite lors des operations de chargement d'un reacteur nucleaire REP

    Verdier, A


    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-{alpha} and Feynman-{alpha} methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  20. Effect of sensitization and cold work on stress corrosion susceptibility of austenitic stainless steels in boiling water reactor (BWR) and pressurized water reactor (PWR) conditions

    Haenninen, H.; Aho-Mantila, I.


    The influence of metallurgical variables on stress corrosion cracking of austenitic stainless steels, in particular AISI 304 and OX18H10T, was examined in O/sub 2/ enriched BWR conditions (8 ppm O/sub 2/) and in typical PWR conditions. Cracking susceptibility in BWR conditions is especially sensitive to alpha martensite content and sensitization. Cracking in alpha martensite compounds is intergranular and transgranular and it can not be related to sensitization. Sensitization induces cracking only in creviced conditions (double U-bend specimens) in AISI 304 steels. In creviced conditions OX18H10T steel exhibits cracking in solution annealed, stabilized and sensitized conditions. The sensitized material is most susceptible. Cracking in solution annealed and stabilized OX18H10T steel is intergranular and transgranular. In PWR conditions (O/sub 2/ content 2 ppb) no cracking is observed. (ESA)

  1. An On-line Assessment System for English-Chinese Translation

    TIAN Yan; LU Ru-zhan; DUAN Jian-yong


    On-line assessment of English-Chinese translation is a challenging task as it involves natural language processing. YanFa, an on-line assessment system for English-Chinese translation, is a pilot research project into scoring student's translation on-line. Based on the theory of translation equivalence, an algorithm called "conceptual similarity matching" was developed. YanFa can assess students' translation on-line timely, generate test papers automatically, offer standard versions of translation, and the scores of each sentence to students. The evaluation proves that YanFa is practical compared with the scores given by experts.

  2. On-Line Booking Policies and Competitive Analysis of Medical Examination in Hospital

    Li Luo


    Full Text Available From the on-line point, we consider the hospital’s medical examination appointment problem with hierarchical machines. This approach eliminates the need for both demand forecasts and a risk-neutrality assumption. Due to different unit revenue, uncertain demand, and arrival of patients, we design on-line booking policies for two kinds of different situations from the perspective of on-line policy and competitive analysis. After that, we prove the optimal competitive ratios. Through numerical examples, we compare advantages and disadvantages between on-line policies and traditional policies, finding that there is different superiority for these two policies under different arrival sequences.


    Reviewed by Dr. Abdullah KUZU


    Full Text Available ON-LINE EDUCATION: AN EMANCIPATING VISION Written by Margarita Victoria Gomez Publisher: Cortez Editora ISBN 85-249-1062-3 São Paulo, Brazil, July 2004. Reviewed by Ricardo Romo TORRES Manuel Moreno Castaneda Maria del Sol Orozco Aguirre Universidad de Guadalajara Virtual MEXICO "On-line Education " is a book that meets the challenge to present itself as an unfinished text. In addition to demanding an active participation of the reader, it requires a predisposition to complete it through an also inconclusive dialogue. Among its goals we can find the search for education networks that remain under constant reconstruction, maintaining the emancipative vision as an imperative for the reconfiguration, as a constituted tissue for the scaffolding of subjects who impress,imagine, think and have a will. The unfinished part is presented in the following paragraph: "Similar to Babel according to Borges, knowledge is the product of a hazard in which fiction is the universe and where, out of fear of its multiple combinations, individual texts constitute an unfinished tale. In this labyrinth-shaped universe, the mirror and the recurrence do not allow finding the way out" page 134. The sense of being of man within a planetary community is framed by the ontological condition of conclusiveness; the requirement of being more, however, as well. This more is guided by the need for dialogue with others in terms of the opening and maintaining an unconcluded dialogue. The idea of a web or network leads to an articulating demand within the context of the category of wholeness. Based on this category, we will be able to present a reticulum of concepts where we can find included the web, subjectivity, identity, experience, mediation and, of course, digital literacy. This reticulum results in a concept of identified education with a condensed arrival point to something that the author defines as 'the pedagogy of virtuality'. The author is

  4. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter [NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany); Bertholdt, Horst-Otto [NCT Consulting, Leonhardstrasse 16-18, 90443 Nuernberg (Germany); Adams, Andreas; Impertro, Michael; Roesch, Josef [RWE Power, 68643 Biblis (Germany)


    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  5. Methodology to evaluate the crack growth rate by stress corrosion cracking in dissimilar metals weld in simulated environment of PWR nuclear reactor

    Paula, Raphael G.; Figueiredo, Celia A.; Rabelo, Emerson G., E-mail:, E-mail:, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    Inconel alloys weld metal is widely used to join dissimilar metals in nuclear reactors applications. It was recently observed failures of weld components in plants, which have triggered an international effort to determine reliable data on the stress corrosion cracking behavior of this material in reactor environment. The objective of this work is to develop a methodology to determine the crack growth rate caused by stress corrosion in Inconel alloy 182, using the specimen (Compact Tensile) in simulated PWR environment. (author)

  6. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    James, L.A.


    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  7. Experiment on the improvement of sinterability for dry recycling nuclear fuel pellets by using simulated spent PWR fuel of high burnup

    Kim, Woong Ki; Kim, S. S.; Park, G. I.; Lee, Jae W.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Lee, J. W.; Yang, M. S.; Shin, W. C


    To study the fabrication characteristics of dry recycling nuclear fuel using spent PWR fuel with high burnup of 60,000 MWd/tU, the fission products of spent PWR fuel was analyzed by ORIGEN-2 code. Simulated spent PWR fuel pellets were fabricated by using UO{sub 2} powder added by the simulated fission products. The simulated dry-recycling-fuel pellets were fabricated by dry recycling fuel fabrication flow including 3 cycle treated OREOX(Oxidation and REduction of OXide fuel) process. A small amount of dopant such as TiO{sub 2}, Nb{sub 2}O{sub 5}, Li{sub 2}O are added to increase sinterability of the OREOX treated powder. As the results of experiments, the densities of sintered pellets without dopant ranged from 10.04 to 10.34 g/cm{sup 3}(94.3 to 97.1% of T.D.), the grain size of the pellets ranged from 3 to 4 {mu}m. The sintered density of the pellets with TiO{sub 2} or Nb{sub 2}O{sub 5} ranged from 10.46 to 10.32 g/cm{sup 3}(98.2 to 96.9 % of T.D.) The grain size of the pellets with TiO{sub 2}, Nb{sub 2}O{sub 5} or Li{sub 2}O ranged from 7.3 to 12.2 {mu}m.

  8. Acceptance test for 900 MWe PWR unit replacement steam generators; Essai de reception des generateurs de vapeur de remplacement des tranches REP 900

    Gourguechon, B.


    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG`s differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs.

  9. Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant

    Yuann, Yng-Ruey, E-mail:; Hsu, Keng-Hsien, E-mail:; Lin, Chin-Tsu, E-mail:


    Highlights: • Calculate NSSS and containment transient response during extended SBO of 24 h. • RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. • Reactor coolant pump seal leakage is specifically modeled for each loop. • Analyses are performed with and without secondary-side depressurization, respectively. • Considering different total available time for turbine driven auxiliary feedwater system. - Abstract: A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP5-3D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS

  10. Enhancing On-Line Teaching with Verbal Immediacy through Self-Determination Theory

    Furlich, Stephen A.


    This paper explores the use of instructor verbal immediacy behaviors for on-line classes. Specifically, it demonstrates how instructor verbal immediacy behaviors found in face-to-face classes can also be displayed for on-line classes. It is argued that self-determination theory describes identification of the student as an important role in the…

  11. A novel, optical, on-line bacteria sensor for monitoring drinking water quality

    Højris, Bo; Christensen, Sarah Christine Boesgaard; Albrechtsen, Hans-Jørgen;


    Today, microbial drinking water quality is monitored through either time-consuming laboratory methods or indirect on-line measurements. Results are thus either delayed or insufficient to support proactive action. A novel, optical, on-line bacteria sensor with a 10-minute time resolution has been...... conditions such as pollution events in drinking water....

  12. Analyze On-line Star Economy Basing on Models of Entrepreneurship



    The outstanding performance of the On-line Star Economy is bound up with social media and promotion by fans, stimulating a new round of consumption upgrading and capital tendency. There is no denying that the On-line Star Economy may be the fortuitous outcome of the times. But the fact remains it can be analyzed rationally using Models of Entrepreneurship.

  13. On-line monitoring and control of animal-cell cultures

    Pol, van der J.J.


    On-line analysis and control of biotechnological processes is still the stepchild in industry. In general, only parameters as dissolved-oxygen concentration, pH and temperature are controlled on-line. Important parameters as substrate and inhibitor concentrations are only measured

  14. On-Line Multichannel Raman Spectroscopic Detection System For Capillary Zone Electrophoresis


    An on-line multichannel Raman spectroscopic detection system for capillary electrophoresis was established by using an Ar+ laser and a cryogenically cooled ICCD. Resonant excitation Raman spectra of methyl red and methyl orange were employed to test the system. The result shows that it could yield on-line electrophoretogram and time series of Raman spectra.

  15. High School Open On-Line Courses (HOOC): A Case Study from Italy

    Canessa, Enrique; Pisani, Armando


    The first implementation of complete high school, open on-line courses (HOOC) aiming to support the training and basic scientific knowledge of young students from the Liceo Ginnasio Dante Alighieri in Gorizia, Italy, is discussed. Using the open source and automated recording system openEyA, HOOC give a student the opportunity to watch on-line, at…

  16. Challenges and opportunities in ‘last mile’ logistics for on-line food retail

    Trienekens, Jacques; Hvolby, Hans Henrik; Turner, Paul


    Conventional approaches to logistics for food retail continue to be challenged by the rapid growth of on-line food retail. At the same time, ‘last mile’ logistics optimization for on-line retail also face challenges as changing consumer expectations, habits and purchasing patterns intersect with

  17. A New Method of On-line Grid Impedance Estimation for PV Inverter

    Teodorescu, Remus; Asiminoaei, Lucian; Blaabjerg, Frede


    for on-line measuring the grid impedance is presented. The presented method requires no extra hardware being accommodated by typical PV inverters, sensors and CPU, to provide a fast and low cost approach of on-line impedance measurement. By injecting a non-characteristic harmonic current and measuring...

  18. The Anatomy of Program Design for an On-Line Business Management Course

    Barger, Bonita


    How does one design an on-line course to bridge theory and practice? How can the feedback of on-going stakeholder (student and administration) be incorporated into the design process to enhance quality? This paper presents the theoretical underpinning of designing an on-line management course recognized as best practice for a "well organized…

  19. High-temperature compatibility between liquid metal as PWR fuel gap filler and stainless steel and high-density concrete

    Wongsawaeng, Doonyapong; Jumpee, Chayanit; Jitpukdee, Manit


    In conventional nuclear fuel rods for light-water reactors, a helium-filled as-fabricated gap between the fuel and the cladding inner surface accommodates fuel swelling and cladding creep down. Because helium exhibits a very low thermal conductivity, it results in a large temperature rise in the gap. Liquid metal (LM; 1/3 weight portion each of lead, tin, and bismuth) has been proposed to be a gap filler because of its high thermal conductivity (∼100 times that of He), low melting point (∼100 °C), and lack of chemical reactivity with UO2 and water. With the presence of LM, the temperature drop across the gap is virtually eliminated and the fuel is operated at a lower temperature at the same power output, resulting in safer fuel, delayed fission gas release and prevention of massive secondary hydriding. During normal reactor operation, should an LM-bonded fuel rod failure occurs resulting in a discharge of liquid metal into the bottom of the reactor pressure vessel, it should not corrode stainless steel. An experiment was conducted to confirm that at 315 °C, LM in contact with 304 stainless steel in the PWR water chemistry environment for up to 30 days resulted in no observable corrosion. Moreover, during a hypothetical core-melt accident assuming that the liquid metal with elevated temperature between 1000 and 1600 °C is spread on a high-density concrete basement of the power plant, a small-scale experiment was performed to demonstrate that the LM-concrete interaction at 1000 °C for as long as 12 h resulted in no penetration. At 1200 °C for 5 h, the LM penetrated a distance of ∼1.3 cm, but the penetration appeared to stop. At 1400 °C the penetration rate was ∼0.7 cm/h. At 1600 °C, the penetration rate was ∼17 cm/h. No corrosion based on chemical reactions with high-density concrete occurred, and, hence, the only physical interaction between high-temperature LM and high-density concrete was from tiny cracks generated from thermal stress. Moreover

  20. Improvement plans for the RHIC/AGS on-line model environments

    Brown,K.A.; Ahrens, L.; Beebe-Wang, J.; Morris, J.; Nemesure, S.; Robert-Demolaize, G.; Satogata, T.; Schoefer, V.; Tepikian, S.


    The on-line models for Relativistic Ion Collider (RHIC) and the RHIC pre-injectors (the AGS and the AGS Booster) can be thought of as containing our best collective knowledge of these accelerators. As we improve these on-line models we are building the framework to have a sophisticated model-based controls system. Currently the RHIC on-line model is an integral part of the controls system, providing the interface for tune control, chromaticity control, and non-linear chromaticity control. What we discuss in this paper is our vision of the future of the on-line model environment for RHIC and the RHIC preinjectors. Although these on-line models are primarily used as Courant-Snyder parameter calculators using live machine settings, we envision expanding these environments to encompass many other problem domains.