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Sample records for low-oxygen hydrogen-rich coolants

  1. Coal pyrolysis under hydrogen-rich gases

    Energy Technology Data Exchange (ETDEWEB)

    Liao, H.; Sun, C.; Li, B.; Liu, Z. [Chinese Academy of Sciences, Taiyuan (China). State Key Laboratory of Coal Conversion, Institute of Coal Chemistry

    1998-04-01

    To improve the economy of the pyrolysis process by reducing the hydrogen cost, it is suggested to use cheaper hydrogen-rich gases such as coke-oven gas (COG) or synthesis gas (SG) instead of pure hydrogen. The pyrolysis of Chinese Xianfeng lignite which was carried out with real COG and SG at 3-5 MPa, a final temperature of 650{degree}C and a heating rate of 5{degree}C/min in a 10g fixed-bed reactor is compared with coal pyrolysis with pure hydrogen and nitrogen under the same conditions. The results indicate that compared with hydropyrolysis at the same total pressure, the total conversion and tar yields from coal pyrolysis with COG and SG decreases while the unwanted water increases. However, at the same H{sub 2} partial pressure, the tar yields and yields of BBTX, PCX and naphthalene from the pyrolysis of coal with COG and SG are all significantly higher than those of hydropyrolysis. Therefore, it is possible to use COG and SG instead of pure hydrogen. 8 refs., 3 figs., 6 tabs.

  2. Thermonuclear runaways in thick hydrogen rich envelopes of neutron stars

    Science.gov (United States)

    Starrfield, S. G.; Kenyon, S.; Truran, J. W.; Sparks, W. M.

    1981-01-01

    A Lagrangian, fully implicit, one dimensional hydrodynamic computer code was used to evolve thermonuclear runaways in the accreted hydrogen rich envelopes of 1.0 Msub solar neutron stars with radii of 10 km and 20 km. Simulations produce outbursts which last from about 750 seconds to about one week. Peak effective temeratures and luninosities were 26 million K and 80 thousand Lsub solar for the 10 km study and 5.3 millison and 600 Lsub solar for the 20 km study. Hydrodynamic expansion on the 10 km neutron star produced a precursor lasting about one ten thousandth seconds.

  3. Stable Hydrogen-rich Atmospheres of Young Rocky Planets

    Science.gov (United States)

    Zahnle, K. J.; Catling, D. C.; Gacesa, M.

    2016-12-01

    SourceURL:file://localhost/Volumes/Lexar/Zahnle_AGU_2016.docx Understanding hydrogen escape is essential to understanding the limits to habitability, both for liquid water where the Sun is bright, but also to assess the true potential of H2 as a greenhouse gas where the Sun is faint. Hydrogen-rich primary atmospheres of Earth-like planets can result either from gravitational capture of solar nebular gases (with helium), or from impact shock processing of a wide variety of volatile-rich planetesimals (typically accompanied by H2O, CO2, and under the right circumstances, CH4). Most studies of hydrogen escape from planets focus on determining how fast the hydrogen escapes. In general this requires solving hydrodynamic equations that take into account the acceleration of hydrogen through a critical transonic point and an energy budget that should include radiative heating and cooling, thermal conduction, the work done in lifting the hydrogen against gravity, and the residual heat carried by the hydrogen as it leaves. But for planets from which hydrogen escape is modest or insignificant, the atmosphere can be approximated as hydrostatic, which is much simpler, and for which a relatively full-featured treatment of radiative cooling by embedded molecules, atoms, and ions such as CO2 and H3+ is straightforward. Previous work has overlooked the fact that the H2 molecule is extremely efficient at exciting non-LTE CO2 15 micron emission, and thus that radiative cooling can be markedly more efficient when H2 is abundant. We map out the region of phase space in which terrestrial planets keep hydrogen-rich atmospheres, which is what we actually want to know for habitability. We will use this framework to reassess Tian et al's (Science 308, pp. 1014-1017, 2005) hypothesis that H2-rich atmospheres may have been rather long-lived on Earth itself. Finally, we will address the empirical observation that rocky planets with thin or negligible atmospheres are rarely or never bigger than

  4. Structure and stabilization of hydrogen-rich transverse.

    Energy Technology Data Exchange (ETDEWEB)

    Lyra, Sgouria [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Wilde, B [Georgia Inst. of Technology, Atlanta, GA (United States); Kolla, Hemanth [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Seitzman, J. [Georgia Inst. of Technology, Atlanta, GA (United States); Lieuwen, T. C. [Georgia Inst. of Technology, Atlanta, GA (United States); Chen, Jacqueline H. [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2014-07-01

    This paper reports the results of a joint experimental and numerical study of the ow characteristics and flame stabilization of a hydrogen rich jet injected normal to a turbulent, vitiated cross ow of lean methane combustion products. Simultaneous high-speed stereoscopic PIV and OH PLIF measurements were obtained and analyzed alongside three-dimensional direct numerical simulations of inert and reacting JICF with detailed H2/CO chemistry. Both the experiment and the simulation reveal that, contrary to most previous studies of reacting JICF stabilized in low-to-moderate temperature air cross ow, the present conditions lead to an autoigniting, burner-attached flame that initiates uniformly around the burner edge. Significant asymmetry is observed, however, between the reaction zones located on the windward and leeward sides of the jet, due to the substantially different scalar dissipation rates. The windward reaction zone is much thinner in the near field, while also exhibiting significantly higher local and global heat release than the much broader reaction zone found on the leeward side of the jet. The unsteady dynamics of the windward shear layer, which largely control the important jet/cross flow mixing processes in that region, are explored in order to elucidate the important flow stability implications arising in the reacting JICF. Vorticity spectra extracted from the windward shear layer reveal that the reacting jet is globally unstable and features two high frequency peaks, including a fundamental mode whose Strouhal number of ~0.7 agrees well with previous non-reacting JICF stability studies. The paper concludes with an analysis of the ignition, ame stabilization, and global structure of the burner-attached flame. Chemical explosive mode analysis (CEMA) shows that the entire windward shear layer, and a large region on the leeward side of the jet, are highly explosive prior to ignition and are dominated by non-premixed flame structures after

  5. Effects of hydrogen rich water on prolonged intermittent exercise.

    Science.gov (United States)

    Da Ponte, Alessandro; Giovanelli, Nicola; Nigris, Daniele; Lazzer, Stefano

    2018-05-01

    Recent studies showed a positive effect of hydrogen rich water (HRW) intake on acid-base homeostasis at rest. We investigated 2-weeks of HRW intake on repeated sprint performance and acid-base status during prolonged intermittent cycling exercise. In a cross over single-blind protocol, 8 trained male cyclists (age [mean±SD] 41±7 years, body mass 72.3±4.4 kg, height 1.77±0.04 m, maximal oxygen uptake [V̇O2max] 52.6±4.4 mL·kg-1·min-1) were provided daily with 2 liters of placebo normal water (PLA, pH 7.6, oxidation/reduction potential [ORP] +230 mV, free hydrogen content 0 ppb) or HRW (pH 9.8, ORP -180 mV, free Hydrogen 450 ppb). Tests were performed at baseline and after each period of 2 weeks of treatment. The treatments were counter-balanced and the sequence randomized. The 30-minute intermittent cycling trial consisted in 10 3-minute blocks, each one composed by 90 seconds at 40% V̇O2max, 60 seconds at 60% V̇O2max, 16 seconds all out sprint, and 14 seconds active recovery. Oxygen uptake (V̇O2), heart rate and power output were measured during the whole test, while mean and peak power output (PPO), time to peak power and Fatigue Index (FI) were determined during all the 16 seconds sprints. Lactate, pH and bicarbonate (HCO3-) concentrations were determined at rest and after each sprint on blood obtained by an antecubital vein indwelling catheter. In the PLA group, PPO in absolute values decreased significantly at the 8th and 9th of 10 sprints and in relative values, ΔPPO, decreased significantly at 6th, 8th and 9th of 10 sprints (by mean: -12±5%, Pmay help to maintain PPO in repetitive sprints to exhaustion over 30 minutes.

  6. Effects of Hydrogen-Rich Saline on Hepatectomy-Induced Postoperative Cognitive Dysfunction in Old Mice.

    Science.gov (United States)

    Tian, Yue; Guo, Shanbin; Zhang, Yan; Xu, Ying; Zhao, Ping; Zhao, Xiaochun

    2017-05-01

    This study aims to investigate the protective effects and underlying mechanisms of hydrogen-rich saline on the cognitive functions of elder mice with partial hepatectomy-induced postoperative cognitive dysfunction (POCD). Ninety-six old male Kunming mice were randomly divided into 4 groups (n = 24 each): control group (group C), hydrogen-rich saline group (group H), POCD group (group P), and POCD + hydrogen-rich saline group (group PH). Cognitive function was subsequently assessed using Morris water-maze (MWM) test. TNF-α and IL-1β levels were measured by enzyme-linked immunosorbent assay (ELISA) and immunohistochemistry, along with NF-κB activity determined by ELISA. The morphology of hippocampal tissues were further observed by HE staining. Learning and memory abilities of mice were significantly impaired at day 10 and day 14 post-surgery, as partial hepatectomy significantly prolonged the escape latency, decreased time at the original platform quadrant and frequency of crossing in group P when compared to group C (p hydrogen-rich saline (group PH) partially rescued spatial memory and learning as it shortened escape latency and increased time and crossing frequency of original platform compared to group P (p hydrogen-rich saline. Hydrogen-rich saline can alleviate POCD via inhibiting NF-κB activity in the hippocampus and reducing inflammatory response.

  7. Investigation on the production of hydrogen rich gas in a plasma converter for motorcycle applications

    International Nuclear Information System (INIS)

    Horng, R.-F.; Chang, Y.-P.; Wu, S.-C.

    2006-01-01

    A plasma fuel converter producing a hydrogen rich gas fuel has been designed and constructed. The methodology included using a high voltage electric arc generator to ionize the mixture of methane fuel and air, which was then reformed into a hydrogen rich gas. It transpired from the experiment that the higher the arc frequency, the higher was the generated hydrogen concentration, with a maximum concentration of 43 vol.% attained with an arc frequency of 200 Hz and an O/C (O 2 /CH 4 ) ratio of 0.10. The maximum hydrogen yield of 0.55 was obtained with an arc frequency of 200 Hz and an O/C ratio between 0.20 and 0.25. By fueling a four stroke motorcycle engine with the hydrogen rich gas, low emissions during the cold start idle condition can be obtained

  8. Attenuation of cigarette smoke-induced airway mucus production by hydrogen-rich saline in rats.

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    Yunye Ning

    Full Text Available BACKGROUND: Over-production of mucus is an important pathophysiological feature in chronic airway disease such as chronic obstructive pulmonary disease (COPD and asthma. Cigarette smoking (CS is the leading cause of COPD. Oxidative stress plays a key role in CS-induced airway abnormal mucus production. Hydrogen protected cells and tissues against oxidative damage by scavenging hydroxyl radicals. In the present study we investigated the effect of hydrogen on CS-induced mucus production in rats. METHODS: Male Sprague-Dawley rats were divided into four groups: sham control, CS group, hydrogen-rich saline pretreatment group and hydrogen-rich saline control group. Lung morphology and tissue biochemical changes were determined by immunohistochemistry, Alcian Blue/periodic acid-Schiff staining, TUNEL, western blot and realtime RT-PCR. RESULTS: Hydrogen-rich saline pretreatment attenuated CS-induced mucus accumulation in the bronchiolar lumen, goblet cell hyperplasia, muc5ac over-expression and abnormal cell apoptosis in the airway epithelium as well as malondialdehyde increase in the BALF. The phosphorylation of EGFR at Tyr1068 and Nrf2 up-regulation expression in the rat lungs challenged by CS exposure were also abrogated by hydrogen-rich saline. CONCLUSION: Hydrogen-rich saline pretreatment ameliorated CS-induced airway mucus production and airway epithelium damage in rats. The protective role of hydrogen on CS-exposed rat lungs was achieved at least partly by its free radical scavenging ability. This is the first report to demonstrate that intraperitoneal administration of hydrogen-rich saline protected rat airways against CS damage and it could be promising in treating abnormal airway mucus production in COPD.

  9. Hydrogen-rich saline ameliorates the severity of L-arginine-induced acute pancreatitis in rats

    International Nuclear Information System (INIS)

    Chen, Han; Sun, Yan Ping; Li, Yang; Liu, Wen Wu; Xiang, Hong Gang; Fan, Lie Ying; Sun, Qiang; Xu, Xin Yun; Cai, Jian Mei; Ruan, Can Ping; Su, Ning; Yan, Rong Lin; Sun, Xue Jun; Wang, Qiang

    2010-01-01

    Molecular hydrogen, which reacts with the hydroxyl radical, has been considered as a novel antioxidant. Here, we evaluated the protective effects of hydrogen-rich saline on the L-arginine (L-Arg)-induced acute pancreatitis (AP). AP was induced in Sprague-Dawley rats by giving two intraperitoneal injections of L-Arg, each at concentrations of 250 mg/100 g body weight, with an interval of 1 h. Hydrogen-rich saline (>0.6 mM, 6 ml/kg) or saline (6 ml/kg) was administered, respectively, via tail vein 15 min after each L-Arg administration. Severity of AP was assessed by analysis of serum amylase activity, pancreatic water content and histology. Samples of pancreas were taken for measuring malondialdehyde and myeloperoxidase. Apoptosis in pancreatic acinar cell was determined with terminal deoxynucleotidyl transferase-mediated deoxyuridine triphosphate nick-end labeling technique (TUNEL). Expression of proliferating cell nuclear antigen (PCNA) and nuclear factor kappa B (NF-κB) were detected with immunohistochemistry. Hydrogen-rich saline treatment significantly attenuated the severity of L-Arg-induced AP by ameliorating the increased serum amylase activity, inhibiting neutrophil infiltration, lipid oxidation and pancreatic tissue edema. Moreover, hydrogen-rich saline treatment could promote acinar cell proliferation, inhibit apoptosis and NF-κB activation. These results indicate that hydrogen treatment has a protective effect against AP, and the effect is possibly due to its ability to inhibit oxidative stress, apoptosis, NF-κB activation and to promote acinar cell proliferation.

  10. Hyperoxygenated hydrogen-rich solution suppresses shock- and resuscitation-induced liver injury.

    Science.gov (United States)

    Dang, Yangjie; Liu, Ting; Mei, Xiaopeng; Meng, Xiangzhong; Gou, Xingchun; Deng, Bin; Xu, Hao; Xu, Lixian

    2017-12-01

    It is not known whether simultaneous delivery of hydrogen and oxygen can reduce injury caused by hemorrhagic shock and resuscitation (HSR). This study investigated the therapeutic potential of hyperoxygenated hydrogen-rich solution (HHOS), a combined hydrogen/oxygen carrier, in a rat model of HSR-induced liver injury. Rats (n = 60) were randomly divided into 5 groups (n = 6 per group at each time point). One group underwent sham operation, and the others were subjected to severe hemorrhagic shock and then treated with lactated Ringer's solution (LRS), hydrogen-rich solution, hyperoxygenated solution, or HHOS. At 2 and 6 h after resuscitation, blood samples (n = 6) were collected from the femoral artery and serum concentrations of alanine aminotransferase and aspartate aminotransferase (AST) were measured. Rats were then sacrificed, and histopathological changes in the liver were evaluated by quantifying the percentage of apoptotic cells by caspase-3 immunohistochemistry and terminal deoxynucleotidyl transferase dUTP nick-end labeling. Inflammation was assessed by assessing malondialdehyde content and tumor necrosis factor-α, and interleukin (IL)-6 expression. Compared to lactated Ringer's solution, hydrogen-rich solution, or hyperoxygenated solution groups, serum AST and alanine aminotransferase levels and IL-6, tumor necrosis factor-α, and malondialdehyde expression in liver tissue were decreased by HHOS treatment. The number of caspase-3- and terminal deoxynucleotidyl transferase dUTP nick end labeling-positive cells was decreased (P < 0.05) by HHOS treatment, 2 and 6 h after resuscitation. HHOS has protective effects against liver injury in a rat model of HSR. Copyright © 2017 Elsevier Inc. All rights reserved.

  11. Therapeutic Effects of Hydrogen-Rich Solution on Aplastic Anemia in Vivo

    Directory of Open Access Journals (Sweden)

    Sanhu Zhao

    2013-08-01

    Full Text Available Background: Aplasitc anemia (AA is a bone marrow failure syndrome characterized by an immune-mediated destruction of hematopoietic stem cells. Though clinical symptoms could be ameliorated by bone marrow transplantation and/or immunosuppressive therapy, frequent recurrence and especially evolution of clonal hematologic diseases remains problematic clinically. Cytokines such as interferon-γ (INF-γ, tumor necrosis factor-α (TNF-α and interleukin-6 (IL-6 secreted by autologous T cells are closely related with the development of AA. Hydrogen-rich solution was reported to inhibit the levels of cytokines including INF-γ, TNF-α and IL-6 in vivo in recent studies. This study was to investigate the potential therapeutic effects of hydrogen-rich solution on AA in vivo. Methods: AA model was determined in vivo by mice and body weights of the mice were used as the basic physiological index. Peripheral blood cells were calculated to evaluate the hematologic recovery degree. Bone marrow nucleated cells (BMNCs, tissue histology, as well as CFU-S and CFU-GM forming units were used to evaluate the recovery of bone marrow microenvironment. The ratio of CD4+ and CD8+ cells were examined along with cytokine levels in serum to determine the efficacy of H2-rich solution on the affected immunological functions. Results: Body weight and number of peripheral blood cells were significantly improved for mice in the H2-rich solution treated groups as compared with those with AA. The number of BMNCs and CFUs increased markedly and the bone marrow microenvironment was also improved significantly. The experimental group restrained the cell apoptosis, relieved hyperemia and accelerated tissue repair. The number of CD4+ and CD8+ cells as well as the ratio of CD4/CD8 increased to normal gradually, while the levels of TNF-α, IFN-γ, and IL-6 in serum decreased after H2-rich solution treatment. Conclusion: Our study firstly showed that hydrogen-rich solution accelerated the

  12. Effect of bioleaching on hydrogen-rich gas production by steam gasification of sewage sludge

    International Nuclear Information System (INIS)

    Li, Hanhui; Chen, Zhihua; Huo, Chan; Hu, Mian; Guo, Dabin; Xiao, Bo

    2015-01-01

    Highlights: • Bioleaching can modify the physicochemical property of sewage sludge. • The enhancement is mainly hydrogen. • Bioleaching can enhance the gas production in gasification of sewage sludge. • Study provides an insight for future application of bioleached sewage sludge. - Abstract: Effect of bioleaching on hydrogen-rich gas production by steam gasification of sewage sludge was carried out in a lab-scale fixed-bed reactor. The influence of sewage sludge solids concentrations (6–14% (w/v) in 2% increments) during the bioleaching process and reactor temperature (600–900 °C in 100 °C increments) on gasification product yields and gas composition were studied. Characterization of samples showed that bioleaching treatment, especially in 6% (w/v) sludge solids concentration, led to metal removal effectively and modifications in the physicochemical property of sewage sludge which was favored for gasification. The maximum gas yield (49.4%) and hydrogen content (46.4%) were obtained at 6% (w/v) sludge solids concentration and reactor temperature of 900 °C. Sewage sludge after the bioleaching treatment may be a feasible feedstock for hydrogen-rich gas product.

  13. Hydrogen-rich saline injection into the subarachnoid cavity within 2 weeks promotes recovery after acute spinal cord injury

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    Jian-long Wang

    2015-01-01

    Full Text Available Hydrogen can relieve tissue-damaging oxidative stress, inflammation and apoptosis. Injection of hydrogen-rich saline is an effective method for transporting molecular hydrogen. We hypothesized that hydrogen-rich saline would promote the repair of spinal cord injury induced by Allen′s method in rats. At 0.5, 1, 2, 4, 8, 12 and 24 hours after injury, then once daily for 2 weeks, 0.25 mL/kg hydrogen-rich saline was infused into the subarachnoid space through a catheter. Results at 24 hours, 48 hours, 1 week and 2 weeks after injury showed that hydrogen-rich saline markedly reduced cell death, inflammatory cell infiltration, serum malondialdehyde content, and caspase-3 immunoreactivity, elevated serum superoxide dismutase activity and calcitonin gene-related peptide immunoreactivity, and improved motor function in the hindlimb. The present study confirms that hydrogen-rich saline injected within 2 weeks of injury effectively contributes to the repair of spinal cord injury in the acute stage.

  14. Effectiveness of hydrogen rich water on antioxidant status of subjects with potential metabolic syndrome-an open label pilot study.

    Science.gov (United States)

    Nakao, Atsunori; Toyoda, Yoshiya; Sharma, Prachi; Evans, Malkanthi; Guthrie, Najla

    2010-03-01

    Metabolic syndrome is characterized by cardiometabolic risk factors that include obesity, insulin resistance, hypertension and dyslipidemia. Oxidative stress is known to play a major role in the pathogenesis of metabolic syndrome. The objective of this study was to examine the effectiveness of hydrogen rich water (1.5-2 L/day) in an open label, 8-week study on 20 subjects with potential metabolic syndrome. Hydrogen rich water was produced, by placing a metallic magnesium stick into drinking water (hydrogen concentration; 0.55-0.65 mM), by the following chemical reaction; Mg + 2H(2)O --> Mg (OH)(2) + H(2). The consumption of hydrogen rich water for 8 weeks resulted in a 39% increase (pfasting glucose levels during the 8 week study. In conclusion, drinking hydrogen rich water represents a potentially novel therapeutic and preventive strategy for metabolic syndrome. The portable magnesium stick was a safe, easy and effective method of delivering hydrogen rich water for daily consumption by participants in the study.

  15. Protective Effects of Hydrogen-Rich Saline Against Lipopolysaccharide-Induced Alveolar Epithelial-to-Mesenchymal Transition and Pulmonary Fibrosis.

    Science.gov (United States)

    Dong, Wen-Wen; Zhang, Yun-Qian; Zhu, Xiao-Yan; Mao, Yan-Fei; Sun, Xue-Jun; Liu, Yu-Jian; Jiang, Lai

    2017-05-19

    BACKGROUND Fibrotic change is one of the important reasons for the poor prognosis of patients with acute respiratory distress syndrome (ARDS). The present study investigated the effects of hydrogen-rich saline, a selective hydroxyl radical scavenger, on lipopolysaccharide (LPS)-induced pulmonary fibrosis. MATERIAL AND METHODS Male ICR mice were divided randomly into 5 groups: Control, LPS-treated plus vehicle treatment, and LPS-treated plus hydrogen-rich saline (2.5, 5, or 10 ml/kg) treatment. Twenty-eight days later, fibrosis was assessed by determination of collagen deposition, hydroxyproline, and type I collagen levels. Development of epithelial-to-mesenchymal transition (EMT) was identified by examining protein expressions of E-cadherin and α-smooth muscle actin (α-SMA). Transforming growth factor (TGF)-β1 content, total antioxidant capacity (T-AOC), malondialdehyde (MDA) content, catalase (CAT), and superoxide dismutase (SOD) activity were determined. RESULTS Mice exhibited increases in collagen deposition, hydroxyproline, type I collagen contents, and TGF-β1 production in lung tissues after LPS treatment. LPS-induced lung fibrosis was associated with increased expression of α-SMA, as well as decreased expression of E-cadherin. In addition, LPS treatment increased MDA levels but decreased T-AOC, CAT, and SOD activities in lung tissues, indicating that LPS induced pulmonary oxidative stress. Hydrogen-rich saline treatment at doses of 2.5, 5, or 10 ml/kg significantly attenuated LPS-induced pulmonary fibrosis. LPS-induced loss of E-cadherin in lung tissues was largely reversed, whereas the acquisition of α-SMA was dramatically decreased by hydrogen-rich saline treatment. In addition, hydrogen-rich saline treatment significantly attenuated LPS-induced oxidative stress. CONCLUSIONS Hydrogen-rich saline may protect against LPS-induced EMT and pulmonary fibrosis through suppressing oxidative stress.

  16. Hydrogen-rich saline protects retina against glutamate-induced excitotoxic injury in guinea pig.

    Science.gov (United States)

    Wei, Lihua; Ge, Li; Qin, Shucun; Shi, Yunzhi; Du, Changqing; Du, Hui; Liu, Liwei; Yu, Yang; Sun, Xuejun

    2012-01-01

    Molecular hydrogen (H(2)) is an efficient antioxidant that can selectively reduce hydroxyl radicals and inhibit oxidative stress-induced injuries. We investigated the protective effects and mechanism of hydrogen-rich saline in a glutamate-induced retinal injury model. Retinal excitotoxicity was induced in healthy guinea pigs by injecting glutamate into the vitreous cavity. After 30 min, hydrogen-rich saline was injected into the vitreous cavity, the peritoneal cavity or both. Seven days later, the retinal stress response was evaluated by examining the stress biomarkers, inducible nitric-oxide synthase (iNOS) and glucose-regulated protein 78 (GRP78). The impaired glutamate uptake was assessed by the expression of the excitatory amino acid transporter 1(EAAT-1). The retinal histopathological changes were investigated, focusing on the thicknesses of the entire retina and its inner layer, the number of cells in the retinal ganglion cell layer (GCL) and the ultrastructure of the retinal ganglion cells (RGCs) and glial cells. Compared with the glutamate-induced injury group, the hydrogen-rich saline treatment reduced the loss of cells in the GCL and thinning of the retina and attenuated cellular morphological damage. These improvements were greatest in animals that received H(2) injections into both the vitreous and the peritoneal cavities. The hydrogen-rich saline also inhibited the expression of glial fibrillary acidic protein (GFAP) in Müller cells, CD11b in microglia, and iNOS and GRP78 in glial cells. Moreover, the hydrogen-rich saline increased the expression of EAAT-1. In conclusion, the administration of hydrogen-rich saline through the intravitreal or/and intraperitoneal routes could reduce the retinal excitotoxic injury and promote retinal recovery. This result likely occurs by inhibiting the activation of glial cells, decreasing the production of the iNOS and GRP78 and promoting glutamate clearance. Copyright © 2011 Elsevier Ltd. All rights reserved.

  17. (F)UV Spectral Analysis of Hot, Hydrogen-Rich Central Stars of Planetary Nebulae

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    Ziegler, M.; Rauch, T.; Werner, K.; Kruk, J. W.

    2010-11-01

    Metal abundances of CSPNe are not well known although they provide important constraints on AGB nucleosynthesis. We aim to determine metal abundances of two hot, hydrogen-rich CSPNe (namely of A35 and NGC3587, the latter also known as M97 or the Owl Nebula) and to derive Teff and log g precisely from high-resolution, high-S/N (far-) ultraviolet observations obtained with FUSE and HST/STIS. For this purpose, we utilize NLTE model atmospheres calculated with TMAP, the Tübingen Model Atmosphere Package. Due to strong line absorption of the ISM, simultaneous modeling of interstellar features has become a standard tool in our analyses. We present preliminary results, demonstrating the importance of combining stellar and interstellar models, in order to clearly identify and measure the strengths of strategic photospheric lines.

  18. A Review on Preferential Oxidation of Carbon Monoxide in Hydrogen Rich Gases

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    A. Mishra

    2011-05-01

    Full Text Available In this review, recent works on the preferential oxidation of carbon monoxide in hydrogen rich gases for fuel cell applications are summarized. H2 is used as a fuel for polymer-electrolyte membrane fuel cell (PEMFC. It is produced by reforming of natural gas or liquid fuels followed by water gas shift reaction. The produced gas consists of H2, CO, and CO2. In which CO content is around 1%, which is highly poisonous for the Pt anode of the PEMFC so that further removal of CO is needed. Catalytic preferential oxidation of CO (CO-PROX is one of the most suitable methods of purification of H2 because of high CO conversion rate at low temperature range, which is preferable for PEMFC operating conditions. Catalysts used for COPROX are mainly noble metal based; gold based and base metal oxide catalysts among them Copper-Ceria based catalysts are the most appropriate due to its low cost, easy availability and result obtained by these catalysts are comparable with the conventional noble metal catalysts. Copyright © 2011 BCREC UNDIP. All rights reserved(Received: 22nd October 2010, Revised: 12nd January 2011, Accepted: 19th January 2011[How to Cite: A. Mishra, R. Prasad. (2011. A Review on Preferential Oxidation of Carbon Monoxide in Hydrogen Rich Gases. Bulletin of Chemical Reaction Engineering & Catalysis, 6 (1: 1-14. doi:10.9767/bcrec.6.1.191.1-14][How to Link / DOI: http://dx.doi.org/10.9767/bcrec.6.1.191.1-14 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/191] | View in 

  19. The Stability of Hydrogen-Rich Atmospheres of Earth-Like Planets

    Science.gov (United States)

    Zahnle, Kevin

    2016-01-01

    Understanding hydrogen escape is essential to understanding the limits to habitability, both for liquid water where the Sun is bright, but also to assess the true potential of H2 as a greenhouse gas where the Sun is faint. Hydrogen-rich primary atmospheres of Earth-like planets can result either from gravitational capture of solar nebular gases (with helium), or from impact shock processing of a wide variety of volatile-rich planetesimals (typically accompanied by H2O, CO2, and under the right circumstances, CH4). Most studies of hydrogen escape from planets focus on determining how fast the hydrogen escapes. In general this requires solving hydro- dynamic equations that take into account the acceleration of hydrogen through a critical transonic point and an energy budget that should include radiative heating and cooling, thermal conduction, the work done in lifting the hydrogen against gravity, and the residual heat carried by the hydrogen as it leaves. But for planets from which hydrogen escape is modest or insignificant, the atmosphere can be approximated as hydrostatic, which is much simpler, and for which a relatively full-featured treatment of radiative cooling by embedded molecules, atoms, and ions such as CO2 and H3+ is straightforward. Previous work has overlooked the fact that the H2 molecule is extremely efficient at exciting non-LTE CO2 15 micron emission, and thus that radiative cooling can be markedly more efficient when H2 is abundant. We map out the region of phase space in which terrestrial planets keep hydrogen-rich atmospheres, which is what we actually want to know for habitability. We will use this framework to reassess Tian et al's hypothesis that H2-rich atmospheres may have been rather long-lived on Earth itself. Finally, we will address the empirical observation that rocky planets with thin or negligible atmospheres are rarely or never bigger than 1.6 Earth radii.

  20. Experimental Evaluation of SI Engine Operation Supplemented by Hydrogen Rich Gas from a Compact Plasma Boosted Reformer

    International Nuclear Information System (INIS)

    J. B. Green, Jr.; N. Domingo; J. M. E. Storey; R.M. Wagner; J.S. Armfield; L. Bromberg; D. R. Cohn; A. Rabinovich; N. Alexeev

    2000-01-01

    It is well known that hydrogen addition to spark-ignited (SI) engines can reduce exhaust emissions and increase efficiency. Micro plasmatron fuel converters can be used for onboard generation of hydrogen-rich gas by partial oxidation of a wide range of fuels. These plasma-boosted microreformers are compact, rugged, and provide rapid response. With hydrogen supplement to the main fuel, SI engines can run very lean resulting in a large reduction in nitrogen oxides (NO x ) emissions relative to stoichiometric combustion without a catalytic converter. This paper presents experimental results from a microplasmatron fuel converter operating under variable oxygen to carbon ratios. Tests have also been carried out to evaluate the effect of the addition of a microplasmatron fuel converter generated gas in a 1995 2.3-L four-cylinder SI production engine. The tests were performed with and without hydrogen-rich gas produced by the plasma boosted fuel converter with gasoline. A one hundred fold reduction in NO x due to very lean operation was obtained under certain conditions. An advantage of onboard plasma-boosted generation of hydrogen-rich gas is that it is used only when required and can be readily turned on and off. Substantial NO x reduction should also be obtainable by heavy exhaust gas recirculation (EGR) facilitated by use of hydrogen-rich gas with stoichiometric operation

  1. Hydrogen-rich gas as a product of two-stage co-gasification of lignite/waste mixtures

    Czech Academy of Sciences Publication Activity Database

    Straka, Pavel; Bičáková, Olga

    2014-01-01

    Roč. 39, č. 21 (2014), s. 10987-10995 ISSN 0360-3199 Institutional support: RVO:67985891 Keywords : co-gasification * waste plastics * lignite * hydrogen-rich gas Subject RIV: DM - Solid Waste and Recycling Impact factor: 3.313, year: 2014 http://authors.elsevier.com/sd/article/S0360319914014025

  2. SN 2017dio: A Type-Ic Supernova Exploding in a Hydrogen-rich Circumstellar Medium

    Science.gov (United States)

    Kuncarayakti, Hanindyo; Maeda, Keiichi; Ashall, Christopher J.; Prentice, Simon J.; Mattila, Seppo; Kankare, Erkki; Fransson, Claes; Lundqvist, Peter; Pastorello, Andrea; Leloudas, Giorgos; Anderson, Joseph P.; Benetti, Stefano; Bersten, Melina C.; Cappellaro, Enrico; Cartier, Régis; Denneau, Larry; Della Valle, Massimo; Elias-Rosa, Nancy; Folatelli, Gastón; Fraser, Morgan; Galbany, Lluís; Gall, Christa; Gal-Yam, Avishay; Gutiérrez, Claudia P.; Hamanowicz, Aleksandra; Heinze, Ari; Inserra, Cosimo; Kangas, Tuomas; Mazzali, Paolo; Melandri, Andrea; Pignata, Giuliano; Rest, Armin; Reynolds, Thomas; Roy, Rupak; Smartt, Stephen J.; Smith, Ken W.; Sollerman, Jesper; Somero, Auni; Stalder, Brian; Stritzinger, Maximilian; Taddia, Francesco; Tomasella, Lina; Tonry, John; Weiland, Henry; Young, David R.

    2018-02-01

    SN 2017dio shows both spectral characteristics of a type-Ic supernova (SN) and signs of a hydrogen-rich circumstellar medium (CSM). Prominent, narrow emission lines of H and He are superposed on the continuum. Subsequent evolution revealed that the SN ejecta are interacting with the CSM. The initial SN Ic identification was confirmed by removing the CSM interaction component from the spectrum and comparing with known SNe Ic and, reversely, adding a CSM interaction component to the spectra of known SNe Ic and comparing them to SN 2017dio. Excellent agreement was obtained with both procedures, reinforcing the SN Ic classification. The light curve constrains the pre-interaction SN Ic peak absolute magnitude to be around {M}g=-17.6 mag. No evidence of significant extinction is found, ruling out a brighter luminosity required by an SN Ia classification. These pieces of evidence support the view that SN 2017dio is an SN Ic, and therefore the first firm case of an SN Ic with signatures of hydrogen-rich CSM in the early spectrum. The CSM is unlikely to have been shaped by steady-state stellar winds. The mass loss of the progenitor star must have been intense, \\dot{M}∼ 0.02{({ε }{{H}α }/0.01)}-1 ({v}{wind}/500 km s‑1) ({v}{shock}/10,000 km s‑1)‑3 M ⊙ yr‑1, peaking at a few decades before the SN. Such a high mass-loss rate might have been experienced by the progenitor through eruptions or binary stripping. Based on observations made with the NOT, operated by the Nordic Optical Telescope Scientific Association at the Observatorio del Roque de los Muchachos, La Palma, Spain, of the Instituto de Astrofisica de Canarias. This work is based (in part) on observations collected at the European Organisation for Astronomical Research in the Southern Hemisphere, Chile as part of PESSTO, (the Public ESO Spectroscopic Survey for Transient Objects Survey) ESO program 188.D-3003, 191.D-0935, 197.D-1075. Based on observations made with the Liverpool Telescope operated on the

  3. Method of controlling injection of oxygen into hydrogen-rich fuel cell feed stream

    Science.gov (United States)

    Meltser, Mark Alexander; Gutowski, Stanley; Weisbrod, Kirk

    2001-01-01

    A method of operating a H.sub.2 --O.sub.2 fuel cell fueled by hydrogen-rich fuel stream containing CO. The CO content is reduced to acceptable levels by injecting oxygen into the fuel gas stream. The amount of oxygen injected is controlled in relation to the CO content of the fuel gas, by a control strategy that involves (a) determining the CO content of the fuel stream at a first injection rate, (b) increasing the O.sub.2 injection rate, (c) determining the CO content of the stream at the higher injection rate, (d) further increasing the O.sub.2 injection rate if the second measured CO content is lower than the first measured CO content or reducing the O.sub.2 injection rate if the second measured CO content is greater than the first measured CO content, and (e) repeating steps a-d as needed to optimize CO consumption and minimize H.sub.2 consumption.

  4. Hydrogen-rich gas production by cogasification of coal and biomass in an intermittent fluidized bed.

    Science.gov (United States)

    Wang, Li-Qun; Chen, Zhao-Sheng

    2013-01-01

    This paper presents the experimental results of cogasification of coal and biomass in an intermittent fluidized bed reactor, aiming to investigate the influences of operation parameters such as gasification temperature (T), steam to biomass mass ratio (SBMR), and biomass to coal mass ratio (BCMR) on hydrogen-rich (H2-rich) gas production. The results show that H2-rich gas free of N2 dilution is produced and the H2 yield is in the range of 18.25~68.13 g/kg. The increases of T, SBMR, and BCMR are all favorable for promoting the H2 production. Higher temperature contributes to higher CO and H2 contents, as well as H2 yield. The BCMR has a weak influence on gas composition, but the yield and content of H2 increase with BCMR, reaching a peak at the BCMR of 4. The H2 content and yield in the product gas increase with SBMR, whilst the content of CO increases first and then decreases correspondingly. At a typical case, the relative linear sensitivity coefficients of H2 production efficiency to T, SBMR, and BCMR were calculated. The results reveal that the order of the influence of the operation parameters on H2 production efficiency is T > SBMR > BCMR.

  5. Plasma steam reforming of E85 for hydrogen rich gas production

    International Nuclear Information System (INIS)

    Zhu Xinli; Hoang Trung; Lobban, Lance L; Mallinson, Richard G

    2011-01-01

    E85 (85 vol% ethanol and 15 vol% gasoline) is a partly renewable fuel that is increasing in supply availability. Hydrogen production from E85 for fuel cell or internal combustion engine applications is a potential method for reducing CO 2 emissions. Steam reforming of E85 using a nonthermal plasma (pulse corona discharge) reactor has been exploited at low temperature (200-300 0 C) without external heating, diluent gas, oxidant or catalyst in this work. Several operational parameters, including the discharge current, E85 concentration and feed flow rate, have been investigated. The results show that hydrogen rich gases (63-67% H 2 and 22-29% CO, with small amounts of CO 2 , C 2 hydrocarbons and CH 4 ) can be produced by this method. A comparison with ethanol reforming and gasoline reforming under identical conditions has also been made and the behaviour of E85 reforming is found to be close to that of ethanol reforming with slightly higher C 2 hydrocarbons yields.

  6. Hydrogen-rich saline attenuates anxiety-like behaviors in morphine-withdrawn mice.

    Science.gov (United States)

    Wen, Di; Zhao, Peng; Hui, Rongji; Wang, Jian; Shen, Qianchao; Gong, Miao; Guo, Hongyan; Cong, Bin; Ma, Chunling

    2017-05-15

    Hydrogen therapy is a new medical approach for a wide range of diseases. The effects of hydrogen on central nervous system-related diseases have recently become increasingly appreciated, but little is known about whether hydrogen affects the morphine withdrawal process. This study aims to investigate the potential effects of hydrogen-rich saline (HRS) administration on naloxone-precipitated withdrawal symptoms and morphine withdrawal-induced anxiety-like behaviors. Mice received gradually increasing doses (25-100 mg/kg, i.p.) of morphine over 3 days. In the naloxone-precipitated withdrawal procedure, the mice were treated with three HRS (20 μg/kg, i.p.) injections, and naloxone (1 mg/kg, i.p.) was given 30 min after HRS administration. Body weight, jumping behavior and wet-dog shakes were immediately assessed. In the spontaneous withdrawal procedure, the mice were treated with HRS (20 μg/kg, i.p.) every 8-h. Mice underwent naloxone-precipitated or spontaneous withdrawal were tested for anxiety-like behaviors in the elevated plus-maze (EPM) and light/dark box (L/D box) paradigm, respectively. In addition, the levels of plasma corticosterone were measured. We found that HRS administration significantly reduced body weight loss, jumping behavior and wet-dog shakes in mice underwent naloxone-precipitated withdrawal, and attenuated anxiety-like behaviors in the EPM and L/D box tests after naloxone-precipitated withdrawal or a 2-day spontaneous withdrawal period. Hypo-activity or motor impairment after HRS administration was not observed in the locomotion tests. Furthermore, HRS administration significantly decreased the levels of corticosterone in morphine-withdrawn mice. These are the first findings to indicate that hydrogen might ameliorate withdrawal symptoms and exert an anxiolytic-like effect in morphine-withdrawal mice. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Influence of the hydrogen-rich on the furnace thermal efficiency

    International Nuclear Information System (INIS)

    Lee, Chien-Li; Jou, Chih-Ju G.

    2016-01-01

    Highlights: • Iν fixed velocity mixing fuel, the flame length is reduced when adding more hydrogen. • Orange-yellowish brightness decrease with increasing tail gas to hydrocarbon fuel. • Adding hydrogen to hydrocarbon fuel will improve the velocity and stability flame. - Abstract: In this research a full-scale furnace is used to recover the hydrogen-rich tail gas as fuel. Adding hydrogen gas to hydrocarbon fuel will reduce the ignition delay of methane, increase the flame velocity and speed up the relatively slow reaction rate of methane to improve the flame stability. The results show that the flame length and orange-yellowish brightness decrease as the amount of tail gas fuel added to the natural gas increases, because of the lower C/H ratio in the flame. Moreover, at a fixed flow rate of hydrocarbon fuel, the moving length of the burning flame is reduced as the amount of hydrogen increases, and thus the visible flame length becomes shorter. Additionally, burning the mixture of tail gas reduces the pressure and increases the gas rising velocity in the furnace radiation and convective zones compared to burning pure tail gas, and thus the gas temperatures in the convective zone and in the flue are raised. The furnace convective zone temperature and the flue gas temperature are 793.6 °C and 350.7 °C, respectively, for burning the mixture fuel (45 vol. % tail gas + 55 vol. % natural gas) vs. 648.5 °C and 346.3 °C for burning the pure tail gas.

  8. Decoration of carbon nano surfaces with hydrogen and hydrogen rich molecules

    International Nuclear Information System (INIS)

    Zöttl, S.

    2013-01-01

    The use of helium nano droplets as a matrix to investigate different atomic and molecular samples is a well established experimental technique. The unique properties of helium allow for different analytical methods and at the same time provide a stable ambient temperature. Cluster growth inside helium nano droplets can be accomplished by repeatedly doping the droplets with sample particles in a controlled environment. The experimental work represented in this thesis was performed using helium nano droplets to create clusters of fullerenes like C 60 and C 70 . The adsorption properties of these fullerene clusters regarding hydrogen and hydrogen rich molecules have been subject to investigation. The observed results suggest that curved carbon nano surfaces offer higher storage densities than planar graphite surfaces. The use of C 60 as a model carbon nano structure provides a well understood molecule for testing and evaluating computational methods to calculate surface properties of various carbon nano materials. The cost effective storage of hydrogen for mobile applications plays a key role in the development of alternatives to fossil fuels. For that reason, the application of carbon nano materials to store hydrogen by adsorption has attracted much scientific attention lately. The insights gained in the presented thesis contribute to the collective efforts and deliver more refined tools to estimate the adsorption properties of future carbon nano materials. In addition to the aforementioned, a time-of-flight mass spectrometer for educational purpose has been designed and constructed in the framework of my PhD thesis. The instrument is successfully used in various lab courses and information on the setup can be found in the Appendix of this work. (author) [de

  9. Stripped-envelope supernova SN 2004dk is now interacting with hydrogen-rich circumstellar material

    Science.gov (United States)

    Mauerhan, Jon C.; Filippenko, Alexei V.; Zheng, WeiKang; Brink, Thomas; Graham, Melissa L.; Shivvers, Isaac; Clubb, Kelsey

    2018-05-01

    The dominant mechanism and time scales over which stripped-envelope supernovae (SNe) progenitor stars shed their hydrogen envelopes are uncertain. Observations of Type Ib and Ic SNe at late phases could reveal the optical signatures of interaction with distant circumstellar material (CSM) providing important clues on the origin of the necessary pre-SN mass loss. We report deep late-time optical spectroscopy of the Type Ib explosion SN 2004dk 4684 days (13 years) after discovery. We detect strong Hα emission with an intermediate line width of ˜400 km s-1 and luminosity ˜2.5 × 1039 erg s-1, signaling that the SN blast wave has caught up with the hydrogen-rich CSM lost by the progenitor system. The line luminosity is the highest ever reported for a SN at this late stage. Prominent emission features of He I, Fe, and Ca are also detected. The spectral characteristics are consistent with CSM energized by the forward shock, and resemble the late-time spectra of the persistently interacting Type IIn SNe 2005ip and 1988Z. We suggest that the onset of interaction with H-rich CSM was associated with a previously reported radio rebrightening at ˜1700 days. The data indicate that the mode of pre-SN mass loss was a relatively slow dense wind that persisted millennia before the SN, followed by a short-lived Wolf-Rayet phase that preceded core-collapse and created a cavity within an extended distribution of CSM. We also present new spectra of SNe 2014C, PTF11iqb, and 2009ip, all of which also exhibit continued interaction with extended CSM distributions.

  10. Alleviation of cadmium toxicity in Medicago sativa by hydrogen-rich water

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Weiti; Gao, Cunyi; Fang, Peng [College of Life Sciences, Nanjing Agricultural University, Nanjing 210095 (China); Lin, Guoqing [Laboratory Center of Life Sciences, Co. Laboratory of Nanjing Agricultural University and Carl Zeiss Far East, Nanjing Agricultural University, Nanjing 210095 (China); Shen, Wenbiao, E-mail: wbshenh@njau.edu.cn [College of Life Sciences, Nanjing Agricultural University, Nanjing 210095 (China)

    2013-09-15

    Highlights: • HRW can alleviate Cd-induced alfalfa seedling growth inhibition and DNA laddering. • HRW alleviates Cd-induced oxidative stress by activating antioxidant enzymes. • Cd uptake in alfalfa seedling roots was decreased by HRW. • HRW can re-establish glutathione homeostasis under Cd stress. -- Abstract: Hydrogen gas (H{sub 2}) induces plant tolerance to several abiotic stresses, including salinity and paraquat exposure. However, the role of H{sub 2} in cadmium (Cd)-induced stress amelioration is largely unknown. Here, pretreatment with hydrogen-rich water (HRW) was used to characterize physiological roles and molecular mechanisms of H{sub 2} in the alleviation of Cd toxicity in alfalfa plants. Our results showed that the addition of HRW at 10% saturation significantly decreased contents of thiobarbituric acid reactive substances (TBARS) caused by Cd, and inhibited the appearance of Cd toxicity symptoms, including the improvement of root elongation and seedling growth. These responses were related to a significant increase in the total or isozymatic activities of representative antioxidant enzymes, or their corresponding transcripts. In vivo imaging of reactive oxygen species (ROS), and the detection of lipid peroxidation and the loss of plasma membrane integrity provided further evidence for the ability of HRW to improve Cd tolerance significantly, which was consistent with a significant enhancement of the ratio of reduced/oxidized (homo)glutathione ((h)GSH). Additionally, plants pretreated with HRW accumulated less amounts of Cd. Together, this study suggested that the usage of HRW could be an effective approach for Cd detoxification and could be explored in agricultural production systems.

  11. Alleviation of cadmium toxicity in Medicago sativa by hydrogen-rich water

    International Nuclear Information System (INIS)

    Cui, Weiti; Gao, Cunyi; Fang, Peng; Lin, Guoqing; Shen, Wenbiao

    2013-01-01

    Highlights: • HRW can alleviate Cd-induced alfalfa seedling growth inhibition and DNA laddering. • HRW alleviates Cd-induced oxidative stress by activating antioxidant enzymes. • Cd uptake in alfalfa seedling roots was decreased by HRW. • HRW can re-establish glutathione homeostasis under Cd stress. -- Abstract: Hydrogen gas (H 2 ) induces plant tolerance to several abiotic stresses, including salinity and paraquat exposure. However, the role of H 2 in cadmium (Cd)-induced stress amelioration is largely unknown. Here, pretreatment with hydrogen-rich water (HRW) was used to characterize physiological roles and molecular mechanisms of H 2 in the alleviation of Cd toxicity in alfalfa plants. Our results showed that the addition of HRW at 10% saturation significantly decreased contents of thiobarbituric acid reactive substances (TBARS) caused by Cd, and inhibited the appearance of Cd toxicity symptoms, including the improvement of root elongation and seedling growth. These responses were related to a significant increase in the total or isozymatic activities of representative antioxidant enzymes, or their corresponding transcripts. In vivo imaging of reactive oxygen species (ROS), and the detection of lipid peroxidation and the loss of plasma membrane integrity provided further evidence for the ability of HRW to improve Cd tolerance significantly, which was consistent with a significant enhancement of the ratio of reduced/oxidized (homo)glutathione ((h)GSH). Additionally, plants pretreated with HRW accumulated less amounts of Cd. Together, this study suggested that the usage of HRW could be an effective approach for Cd detoxification and could be explored in agricultural production systems

  12. Hydrogen-rich water affected blood alkalinity in physically active men.

    Science.gov (United States)

    Ostojic, Sergej M; Stojanovic, Marko D

    2014-01-01

    Possible appliance of effective and safe alkalizing agent in the treatment of metabolic acidosis could be of particular interest to humans experiencing an increase in plasma acidity, such as exercise-induced acidosis. In the present study we tested the hypothesis that the daily oral intake of 2L of hydrogen-rich water (HRW) for 14 days would increase arterial blood alkalinity at baseline and post-exercise as compared with the placebo. This study was a randomized, double blind, placebo-controlled trial involving 52 presumably healthy physically active male volunteers. Twenty-six participants received HRW and 26 a placebo (tap water) for 14 days. Arterial blood pH, partial pressure for carbon dioxide (pCO2), and bicarbonates were measured at baseline and postexercise at the start (day 0) and at the end of the intervention period (day 14). Intake of HRW significantly increased fasting arterial blood pH by 0.04 (95% confidence interval; 0.01 - 0.08; p < 0.001), and postexercise pH by 0.07 (95% confidence interval; 0.01 - 0.10; p = 0.03) after 14 days of intervention. Fasting bicarbonates were significantly higher in the HRW trial after the administration regimen as compared with the preadministration (30.5 ± 1.9 mEq/L vs. 28.3 ± 2.3 mEq/L; p < 0.0001). No volunteers withdrew before the end of the study, and no participant reported any vexatious side effects of supplementation. These results support the hypothesis that HRW administration is safe and may have an alkalizing effect in young physically active men.

  13. Beneficial effects of hydrogen-rich saline on early burn-wound progression in rats.

    Directory of Open Access Journals (Sweden)

    Song Xue Guo

    Full Text Available Deep burn wounds undergo a dynamic process known as wound progression that results in a deepening and extension of the initial burn area. The zone of stasis is more likely to develop more severe during wound progression in the presence of hypoperfusion. Hydrogen has been reported to alleviate injury triggered by ischaemia/reperfusion and burns in various organs by selectively quenching oxygen free radicals. The aim of this study was to investigate the possible protective effects of hydrogen against early burn-wound progression.Deep-burn models were established through contact with a boiled, rectangular, brass comb for 20 s. Fifty-six Sprague-Dawley rats were randomly divided into sham, burn plus saline, and burn plus hydrogen-rich saline (HS groups with sacrifice and analysis at various time windows (6 h, 24 h, 48 h post burn. Indexes of oxidative stress, apoptosis and autophagy were measured in each group. The zone of stasis was evaluated using immunofluorescence staining, ELISA, and Western blot to explore the underlying effects and mechanisms post burn.The burn-induced increase in malondialdehyde was markedly reduced with HS, while the activities of endogenous antioxidant enzymes were significantly increased. Moreover, HS treatment attenuated increases in apoptosis and autophagy postburn in wounds, according to the TUNEL staining results and the expression analysis of Bax, Bcl-2, caspase-3, Beclin-1 and Atg-5 proteins. Additionally, HS lowered the level of myeloperoxidase and expression of TNF-α, IL-1β, and IL-6 in the zone of stasis while augmenting IL-10. The elevated levels of Akt phosphorylation and NF-κB p65 expression post burn were also downregulated by HS management.Hydrogen can attenuate early wound progression following deep burn injury. The beneficial effect of hydrogen was mediated by attenuating oxidative stress, which inhibited apoptosis and inflammation, and the Akt/NF-κB signalling pathway may be involved in regulating the

  14. Co-pyrolysis of waste tire/coal mixtures for smokeless fuel, maltenes and hydrogen-rich gas production

    Czech Academy of Sciences Publication Activity Database

    Bičáková, Olga; Straka, Pavel

    2016-01-01

    Roč. 116, MAY 15 (2016), s. 203-213 ISSN 0196-8904 Grant - others:OPPK(XE) CZ.2.16/3.1.00/21538 Program:OPPK Institutional support: RVO:67985891 Keywords : waste tires * coal * co-pyrolysis * smokeless fuel * tar * hydrogen -rich gas Subject RIV: DM - Solid Waste and Recycling Impact factor: 5.589, year: 2016 http://www.sciencedirect.com/science/article/pii/S0196890416300991

  15. Solubility of metallic elements in LBE under extra low oxygen potential. JFY2003 joint research report

    International Nuclear Information System (INIS)

    Sano, Hiroyuki; Fujisawa, Toshiharu; Furukawa, Tomohiro; Aoto, Kazumi

    2004-03-01

    Lead-Bismuth eutectic alloy (LBE) has been considered as a prospective coolant for a fast-breeder reactor. However a corrosion of cooling pipe is anticipated when it is used at the similar temperature as sodium coolant. In this study, solubility of major metallic elements in LBE was measured under extra low oxygen potential. The interactive effect of those elements on the solubility was also to be examined. (1) The solubility of oxygen in LBE was measured by the gas equilibrium method (1223 k-1323 K). The standard Gibbs free energy change of oxygen solution reaction and the self-interaction parameter of oxygen in LBE were calculated, respectively. (2) The solubility of iron in LBE was measured by both the gas equilibrium method and the oxide equilibrium method (873 K-1323 K). The standard Gibbs free energy change of iron solution reaction, interaction parameter of oxygen on iron and self-interaction parameter of iron in LBE were calculated, respectively. (3) The interactive effect of iron and oxygen on the solubility in LBE was considered thermodynamically. (4) The solubility of chromium and nickel in LBE were measured under Ar-H 2 atmosphere. (author)

  16. Solubility of metallic elements in LBE under extra low oxygen potential. JFY2001 joint research report

    International Nuclear Information System (INIS)

    Sano, Hiroyuki; Fujisawa, Toshiharu

    2002-03-01

    Lead-Bismuth eutectic alloy (LBE) has been considered as a prospective coolant for a fast-breeder reactor. However a corrosion of cooling pipe is anticipated when it is used at the similar temperature as sodium coolant. In this study, solubility of major metallic elements in LBE is to be measured under extra low oxygen potential. The interactive effect of those elements on the solubility is also to be examined. As a first step, measurements of the solubility of iron in LBE at 673 K were conducted where the partial pressure of oxygen was controlled by using equilibrium between iron and its oxide. Several experimental runs were conducted. But relationship between iron content and oxygen content in LBE could not be defined precisely, because chemical reactions proceeded very slowly at such a low temperature and reliable enough data have not been obtained yet until now. Based on the above results, following subjects were extracted for JFY2002 study. (1) To establish the method of quantitative analysis of oxygen content in LBE. (2) To obtain the solubility data at elevated temperature, then approach to lower temperature. (3) To control the oxygen partial pressure in LBE by CO-CO 2 mixed gases supply. (author)

  17. Oral intake of hydrogen-rich water ameliorated chlorpyrifos-induced neurotoxicity in rats

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Tingting; Zhao, Ling; Liu, Mengyu; Xie, Fei; Ma, Xuemei, E-mail: xmma@bjut.edu.cn; Zhao, Pengxiang; Liu, Yunqi; Li, Jiala; Wang, Minglian; Yang, Zhaona; Zhang, Yutong

    2014-10-01

    Chronic exposure to low-levels of organophosphate (OP) compounds, such as chlorpyrifos (CPF), induces oxidative stress and could be related to neurological disorders. Hydrogen has been identified as a novel antioxidant which could selectively scavenge hydroxyl radicals. We explore whether intake of hydrogen-rich water (HRW) can protect Wistar rats from CPF-induced neurotoxicity. Rats were gavaged daily with 6.75 mg/kg body weight (1/20 LD{sub 50}) of CPF and given HRW by oral intake. Nissl staining and electron microscopy results indicated that HRW intake had protective effects on the CPF-induced damage of hippocampal neurons and neuronal mitochondria. Immunostaining results showed that the increased glial fibrillary acidic protein (GFAP) expression in astrocytes induced by CPF exposure can be ameliorated by HRW intake. Moreover, HRW intake also attenuated CPF-induced oxidative stress as evidenced by enhanced level of MDA, accompanied by an increase in GSH level and SOD and CAT activity. Acetylcholinesterase (AChE) activity tests showed significant decrease in brain AChE activity after CPF exposure, and this effect can be ameliorated by HRW intake. An in vitro study demonstrated that AChE activity was more intense in HRW than in normal water with or without chlorpyrifos-oxon (CPO), the metabolically-activated form of CPF. These observations suggest that HRW intake can protect rats from CPF-induced neurotoxicity, and the protective effects of hydrogen may be mediated by regulating the oxidant and antioxidant status of rats. Furthermore, this work defines a novel mechanism of biological activity of hydrogen by directly increasing the AChE activity. - Highlights: • Hydrogen molecules protect rats from CPF-induced damage of hippocampal neurons. • The increased GFAP expression induced by CPF can also be ameliorated by hydrogen. • Hydrogen molecules attenuated the increase in CPF-induced oxidative stress. • Hydrogen molecules attenuated AChE inhibition in vivo

  18. Oral intake of hydrogen-rich water ameliorated chlorpyrifos-induced neurotoxicity in rats

    International Nuclear Information System (INIS)

    Wang, Tingting; Zhao, Ling; Liu, Mengyu; Xie, Fei; Ma, Xuemei; Zhao, Pengxiang; Liu, Yunqi; Li, Jiala; Wang, Minglian; Yang, Zhaona; Zhang, Yutong

    2014-01-01

    Chronic exposure to low-levels of organophosphate (OP) compounds, such as chlorpyrifos (CPF), induces oxidative stress and could be related to neurological disorders. Hydrogen has been identified as a novel antioxidant which could selectively scavenge hydroxyl radicals. We explore whether intake of hydrogen-rich water (HRW) can protect Wistar rats from CPF-induced neurotoxicity. Rats were gavaged daily with 6.75 mg/kg body weight (1/20 LD 50 ) of CPF and given HRW by oral intake. Nissl staining and electron microscopy results indicated that HRW intake had protective effects on the CPF-induced damage of hippocampal neurons and neuronal mitochondria. Immunostaining results showed that the increased glial fibrillary acidic protein (GFAP) expression in astrocytes induced by CPF exposure can be ameliorated by HRW intake. Moreover, HRW intake also attenuated CPF-induced oxidative stress as evidenced by enhanced level of MDA, accompanied by an increase in GSH level and SOD and CAT activity. Acetylcholinesterase (AChE) activity tests showed significant decrease in brain AChE activity after CPF exposure, and this effect can be ameliorated by HRW intake. An in vitro study demonstrated that AChE activity was more intense in HRW than in normal water with or without chlorpyrifos-oxon (CPO), the metabolically-activated form of CPF. These observations suggest that HRW intake can protect rats from CPF-induced neurotoxicity, and the protective effects of hydrogen may be mediated by regulating the oxidant and antioxidant status of rats. Furthermore, this work defines a novel mechanism of biological activity of hydrogen by directly increasing the AChE activity. - Highlights: • Hydrogen molecules protect rats from CPF-induced damage of hippocampal neurons. • The increased GFAP expression induced by CPF can also be ameliorated by hydrogen. • Hydrogen molecules attenuated the increase in CPF-induced oxidative stress. • Hydrogen molecules attenuated AChE inhibition in vivo and in

  19. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  20. Hydrogen-Rich Water Intake Accelerates Oral Palatal Wound Healing via Activation of the Nrf2/Antioxidant Defense Pathways in a Rat Model

    Science.gov (United States)

    Orihuela-Campos, Rita Cristina; Fukui, Makoto; Ito, Hiro-O

    2016-01-01

    The wound healing process attempts to restore the integrity and function of the injured tissue. Additionally, proinflammatory cytokines, growth factors, and oxidative stress play important roles in wound healing. The aim of this study was to determine whether hydrogen-rich water intake induces the activation of the Nrf2/antioxidant defense pathway in rat palatal tissue, thereby reducing systemic oxidative stress and proinflammatory cytokine levels and promoting healing-associated genes. A circular excisional wound was created in the oral palatal region, and the wound healing process was observed. The rats were divided into two experimental groups in which either hydrogen-rich water or distilled water was consumed. In the drinking hydrogen-rich water, the palatal wound healing process was accelerated compared to that in the control group. As molecular hydrogen upregulated the Nrf2 pathway, systemic oxidative stresses were decreased by the activation of antioxidant activity. Furthermore, hydrogen-rich water intake reduced proinflammatory cytokine levels and promoted the expression of healing-associated factors in rat palatal tissue. In conclusion, hydrogen-rich water intake exhibited multiple beneficial effects through activation of the Nrf2/antioxidant defense pathway. The results of this study support the hypothesis that oral administration of hydrogen-rich water benefits the wound healing process by decreasing oxidative stress and inflammatory responses. PMID:26798423

  1. Effects of hydrogen-rich saline on endotoxin-induced uveitis

    Directory of Open Access Journals (Sweden)

    Wei-ming Yan

    2017-01-01

    Full Text Available The therapeutic effects of hydrogen-rich saline (HRS have been reported for a wide range of diseases mainly via selectively reducing the amount of reactive oxygen species. Oxidative stress plays an important role in the pathogenesis of uveitis and endotoxin-induced uveitis (EIU. In this study, we investigated whether HRS can mitigate EIU in rats. Sprague-Dawley rats were randomly divided into Norm group, Model group, HRS group, dexamethasone (DEX group, and rats in the latter three groups were injected with equal amount of lipopolysaccharide (LPS to induce EIU of different severities (by 1 mg/kg of LPS, or 1/8 mg/kg of LPS. Rats in HRS group were injected with HRS intraperitoneally at three different modes to purse an ameliorating effect of EIU (10 mL/kg of HRS immediately after injection of 1 mg/kg of LPS, 20 mL/kg of HRS once a day for 1 week before injection of 1 mg/kg of LPS and at 0, 0.5, 1, 2, 6, 8, 12 hours after LPS administration, or 20 mL/kg of HRS once a day for 1 week before injection of 1/8 mg/kg of LPS, and at 0, 0.5, 1, 2, 6, 8, 12, 24 hours and once a day for 3 weeks after LPS administration. Rats of DEX group were injected with 1 mL/kg of DEX solution intraperitoneally immediately after LPS administration. Rats in Norm and Model groups did not receive any treatment. All rats were examined under slit lamp microscope and graded according to the clinical signs of uveitis. Electroretinogram, quantitative analysis of protein in aqueous humor (AqH and histological examination of iris and ciliary body were also carried out. Our results showed that HRS did not obviously ameliorate the signs of uveitis under slit lamp examination and the inflammatory cells infiltration around iris and cilliary body of EIU induced by 1 mg/kg or 1/8 mg/kg of LPS (P > 0.05, while DEX significantly reduced the inflammation reflected by the above two indicators (P 0.05, while DEX had an obvious therapeutic effect (P < 0.05. However, HRS exerted an inhibition

  2. Hydrogen-rich saline inhibits tobacco smoke-induced chronic obstructive pulmonary disease by alleviating airway inflammation and mucus hypersecretion in rats.

    Science.gov (United States)

    Liu, Zibing; Geng, Wenye; Jiang, Chuanwei; Zhao, Shujun; Liu, Yong; Zhang, Ying; Qin, Shucun; Li, Chenxu; Zhang, Xinfang; Si, Yanhong

    2017-09-01

    Chronic obstructive pulmonary disease induced by tobacco smoke has been regarded as a great health problem worldwide. The purpose of this study is to evaluate the protective effect of hydrogen-rich saline, a novel antioxidant, on chronic obstructive pulmonary disease and explore the underlying mechanism. Sprague-Dawley rats were made chronic obstructive pulmonary disease models via tobacco smoke exposure for 12 weeks and the rats were treated with 10 ml/kg hydrogen-rich saline intraperitoneally during the last 4 weeks. Lung function testing indicated hydrogen-rich saline decreased lung airway resistance and increased lung compliance and the ratio of forced expiratory volume in 0.1 s/forced vital capacity in chronic obstructive pulmonary disease rats. Histological analysis revealed that hydrogen-rich saline alleviated morphological impairments of lung in tobacco smoke-induced chronic obstructive pulmonary disease rats. ELISA assay showed hydrogen-rich saline lowered the levels of pro-inflammatory cytokines (IL-8 and IL-6) and anti-inflammatory cytokine IL-10 in bronchoalveolar lavage fluid and serum of chronic obstructive pulmonary disease rats. The content of malondialdehyde in lung tissue and serum was also determined and the data indicated hydrogen-rich saline suppressed oxidative stress reaction. The protein expressions of mucin MUC5C and aquaporin 5 involved in mucus hypersecretion were analyzed by Western blot and ELISA and the data revealed that hydrogen-rich saline down-regulated MUC5AC level in bronchoalveolar lavage fluid and lung tissue and up-regulated aquaporin 5 level in lung tissue of chronic obstructive pulmonary disease rats. In conclusion, these results suggest that administration of hydrogen-rich saline exhibits significant protective effect on chronic obstructive pulmonary disease through alleviating inflammation, reducing oxidative stress and lessening mucus hypersecretion in tobacco smoke-induced chronic obstructive pulmonary disease rats

  3. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  4. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  5. Reduction of greenhouse gas emission on a medium-pressure boiler using hydrogen-rich fuel control

    International Nuclear Information System (INIS)

    Hsieh, S.-C.; Jou, Chih-Ju G.

    2007-01-01

    The increasing emission of greenhouse gases from the combustion of fossil fuel is believed to be responsible for global warming. A study was carried out to probe the influence of replacing fuel gas with hydrogen-rich refinery gas (R.G.) on the reduction of gas emission (CO 2 and NO x ) and energy saving. Test results show that the emission of CO 2 can be reduced by 16.4% annually (or 21,500 tons per year). The NO x emission can be 8.2% lower, or 75 tons less per year. Furthermore, the use of refinery gas leads to a saving of NT$57 million (approximately US$1.73 million) on fuel costs each year. There are no CO 2 , CO, SO x , unburned hydrocarbon, or particles generated from the combustion of added hydrogen. The hydrogen content in R.G. employed in this study was between 50 and 80 mol%, so the C/H ratio of the feeding fuel was reduced. Therefore, the use of hydrogen-rich fuel has practical benefits for both energy saving and the reduction of greenhouse gas emission

  6. Hydrogen-rich saline may be an effective and specific novel treatment for osteoradionecrosis of the jaw

    Directory of Open Access Journals (Sweden)

    Chen Y

    2015-10-01

    Full Text Available Yuanli Chen, Chunlin Zong, Yuxuan Guo, Lei Tian Department of Cranio-facial Trauma and Orthognathic Surgery Laboratory of Military Stomatology, School of Stomatology, The Fourth Military Medical University, Shaanxi Key Laboratory of Stomatology, Xi’an, People’s Republic of China Abstract: Hydrogen, a therapeutic medical gas, can exert antioxidant activity via selectively reducing cytotoxic reactive oxygen species such as hydroxyl radicals. Hydrogen-rich saline is an alternative form of molecular hydrogen that has been widely used in many studies, including metabolic syndrome, cerebral, hepatic, myocardial ischemia/reperfusion, and liver injuries with obstructive jaundice, with beneficial results. Osteoradionecrosis of the jaw is a serious complication following radiotherapy for head and neck cancers. It has long been known that most radiation-induced symptoms are caused by free radicals generated by radiolysis of H2O, and the hydroxyl radical is the most reactive of these. Reducing the hydroxyl radical can distinctly improve the protection of cells from radiation damage. We hypothesized that hydrogen-rich saline might be an effective and specific method of managing and preventing osteoradionecrosis of the jaw. Keywords: osteoradionecrosis, hydrogen, reactive oxygen species

  7. Preventive Effects of Drinking Hydrogen-Rich Water on Gingival Oxidative Stress and Alveolar Bone Resorption in Rats Fed a High-Fat Diet.

    Science.gov (United States)

    Yoneda, Toshiki; Tomofuji, Takaaki; Kunitomo, Muneyoshi; Ekuni, Daisuke; Irie, Koichiro; Azuma, Tetsuji; Machida, Tatsuya; Miyai, Hisataka; Fujimori, Kouhei; Morita, Manabu

    2017-01-13

    Obesity induces gingival oxidative stress, which is involved in the progression of alveolar bone resorption. The antioxidant effect of hydrogen-rich water may attenuate gingival oxidative stress and prevent alveolar bone resorption in cases of obesity. We examined whether hydrogen-rich water could suppress gingival oxidative stress and alveolar bone resorption in obese rats fed a high-fat diet. Male Fischer 344 rats ( n = 18) were divided into three groups of six rats each: a control group (fed a regular diet and drinking distilled water) and two experimental groups (fed a high-fat diet and drinking distilled water or hydrogen-rich water). The level of 8-hydroxydeoxyguanosine was determined to evaluate oxidative stress. The bone mineral density of the alveolar bone was analyzed by micro-computerized tomography. Obese rats, induced by a high-fat diet, showed a higher gingival level of 8-hydroxydeoxyguanosine and a lower level of alveolar bone density compared to the control group. Drinking hydrogen-rich water suppressed body weight gain, lowered gingival level of 8-hydroxydeoxyguanosine, and reduced alveolar bone resorption in rats on a high-fat diet. The results indicate that hydrogen-rich water could suppress gingival oxidative stress and alveolar bone resorption by limiting obesity.

  8. Inhibition of fungal growth with extreme low oxygen levels

    DEFF Research Database (Denmark)

    Nielsen, Per Væggemose; Haasum, Iben

    1998-01-01

    Fungal spoilage of foods is effectively controlled by removal of oxygen from the package, especially if this is combined with elevated carbon dioxide (CO2) levels. However, great uncertainty exist on just how low the residual oxygen levels in the package must be especially when carbon dioxide lev...... food with low CO2 levels. Active packaging with oxygen absorbers may be considered for these products. The packaging solution must also reflect the micro flora of the product.......Fungal spoilage of foods is effectively controlled by removal of oxygen from the package, especially if this is combined with elevated carbon dioxide (CO2) levels. However, great uncertainty exist on just how low the residual oxygen levels in the package must be especially when carbon dioxide...... Penicillia and Aspergilli were also inhibited by oxygen levels less than 0.5%, but less than 0.01% was required to efficiently inhibit these fungi. Most resistant to very low oxygen levels was the Fusarium species.These results shows that very low oxygen levels are required to avoid fungal growth in package...

  9. Pilot study: Effects of drinking hydrogen-rich water on muscle fatigue caused by acute exercise in elite athletes

    Directory of Open Access Journals (Sweden)

    Aoki Kosuke

    2012-07-01

    Full Text Available Abstract Background Muscle contraction during short intervals of intense exercise causes oxidative stress, which can play a role in the development of overtraining symptoms, including increased fatigue, resulting in muscle microinjury or inflammation. Recently it has been said that hydrogen can function as antioxidant, so we investigated the effect of hydrogen-rich water (HW on oxidative stress and muscle fatigue in response to acute exercise. Methods Ten male soccer players aged 20.9 ± 1.3 years old were subjected to exercise tests and blood sampling. Each subject was examined twice in a crossover double-blind manner; they were given either HW or placebo water (PW for one week intervals. Subjects were requested to use a cycle ergometer at a 75 % maximal oxygen uptake (VO2 for 30 min, followed by measurement of peak torque and muscle activity throughout 100 repetitions of maximal isokinetic knee extension. Oxidative stress markers and creatine kinase in the peripheral blood were sequentially measured. Results Although acute exercise resulted in an increase in blood lactate levels in the subjects given PW, oral intake of HW prevented an elevation of blood lactate during heavy exercise. Peak torque of PW significantly decreased during maximal isokinetic knee extension, suggesting muscle fatigue, but peak torque of HW didn’t decrease at early phase. There was no significant change in blood oxidative injury markers (d-ROMs and BAP or creatine kinease after exercise. Conclusion Adequate hydration with hydrogen-rich water pre-exercise reduced blood lactate levels and improved exercise-induced decline of muscle function. Although further studies to elucidate the exact mechanisms and the benefits are needed to be confirmed in larger series of studies, these preliminary results may suggest that HW may be suitable hydration for athletes.

  10. Ejection of the Massive Hydrogen-rich Envelope Timed with the Collapse of the Stripped SN 2014C

    Energy Technology Data Exchange (ETDEWEB)

    Margutti, Raffaella [Center for Interdisciplinary Exploration and Research in Astrophysics (CIERA), Department of Physics and Astronomy, Northwestern University, Evanston, IL 60208 (United States); Kamble, A.; Milisavljevic, D.; Drout, M.; Chakraborti, S.; Kirshner, R.; Parrent, J. T.; Patnaude, D.; Soderberg, A. M. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Zapartas, E.; De Mink, S. E. [Anton Pannenkoek Institute for Astronomy, University of Amsterdam, 1090 GE Amsterdam (Netherlands); Chornock, R. [Astrophysical Institute, Department of Physics and Astronomy, 251B Clippinger Lab, Ohio University, Athens, OH 45701 (United States); Risaliti, G. [INAF-Arcetri Astrophysical Observatory, Largo E. Fermi 5, I-50125 Firenze (Italy); Zauderer, B. A. [Center for Cosmology and Particle Physics, New York University, 4 Washington Place, New York, NY 10003 (United States); Bietenholz, M. [Department of Physics and Astronomy, York University, Toronto, ON M3J 1P3 (Canada); Cantiello, M. [Kavli Institute for Theoretical Physics, University of California, Santa Barbara, CA 93106 (United States); Chomiuk, L. [Department of Physics and Astronomy, Michigan State University, East Lansing, MI 48824 (United States); Fong, W. [Steward Observatory, University of Arizona, 933 North Cherry Avenue, Tucson, AZ 85721 (United States); Grefenstette, B. [Cahill Center for Astrophysics, 1216 E. California Boulevard, California Institute of Technology, Pasadena, CA 91125 (United States); Guidorzi, C. [University of Ferrara, Department of Physics and Earth Sciences, via Saragat 1, I-44122 Ferrara (Italy); and others

    2017-02-01

    We present multi-wavelength observations of SN 2014C during the first 500 days. These observations represent the first solid detection of a young extragalactic stripped-envelope SN out to high-energy X-rays ∼40 keV. SN 2014C shows ordinary explosion parameters ( E {sub k} ∼ 1.8 × 10{sup 51} erg and M {sub ej} ∼ 1.7 M{sub ⊙}). However, over an ∼1 year timescale, SN 2014C evolved from an ordinary hydrogen-poor supernova into a strongly interacting, hydrogen-rich supernova, violating the traditional classification scheme of type-I versus type-II SNe. Signatures of the SN shock interaction with a dense medium are observed across the spectrum, from radio to hard X-rays, and revealed the presence of a massive shell of ∼1 M {sub ⊙} of hydrogen-rich material at ∼6 × 10{sup 16} cm. The shell was ejected by the progenitor star in the decades to centuries before collapse. This result challenges current theories of massive star evolution, as it requires a physical mechanism responsible for the ejection of the deepest hydrogen layer of H-poor SN progenitors synchronized with the onset of stellar collapse. Theoretical investigations point at binary interactions and/or instabilities during the last nuclear burning stages as potential triggers of the highly time-dependent mass loss. We constrain these scenarios utilizing the sample of 183 SNe Ib/c with public radio observations. Our analysis identifies SN 2014C-like signatures in ∼10% of SNe. This fraction is reasonably consistent with the expectation from the theory of recent envelope ejection due to binary evolution if the ejected material can survive in the close environment for 10{sup 3}–10{sup 4} years. Alternatively, nuclear burning instabilities extending to core C-burning might play a critical role.

  11. Benchmarking DFT and TD-DFT Functionals for the Ground and Excited States of Hydrogen-Rich Peptide Radicals.

    Science.gov (United States)

    Riffet, Vanessa; Jacquemin, Denis; Cauët, Emilie; Frison, Gilles

    2014-08-12

    We assess the pros and cons of a large panel of DFT exchange-correlation functionals for the prediction of the electronic structure of hydrogen-rich peptide radicals formed after electron attachment on a protonated peptide. Indeed, despite its importance in the understanding of the chemical changes associated with the reduction step, the question of the attachment site of an electron and, more generally, of the reduced species formed in the gas phase through electron-induced dissociation (ExD) processes in mass spectrometry is still a matter of debate. For hydrogen-rich peptide radicals in which several positive groups and low-lying π* orbitals can capture the incoming electron in ExD, inclusion of full Hartree-Fock exchange at long-range interelectronic distance is a prerequisite for an accurate description of the electronic states, thereby excluding several popular exchange-correlation functionals, e.g., B3LYP, M06-2X, or CAM-B3LYP. However, we show that this condition is not sufficient by comparing the results obtained with asymptotically correct range-separated hybrids (M11, LC-BLYP, LC-BPW91, ωB97, ωB97X, and ωB97X-D) and with reference CASSCF-MRCI and EOM-CCSD calculations. The attenuation parameter ω significantly tunes the spin density distribution and the excited states vertical energies. The investigated model structures, ranging from methylammonium to hexapeptide, allow us to obtain a description of the nature and energy of the electronic states, depending on (i) the presence of hydrogen bond(s) around the cationic site(s), (ii) the presence of π* molecular orbitals (MOs), and (iii) the selected DFT approach. It turns out that, in the present framework, LC-BLYP and ωB97 yields the most accurate results.

  12. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  13. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  14. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

  15. Aquatic respiration rate measurements at low oxygen concentrations.

    Directory of Open Access Journals (Sweden)

    Moritz Holtappels

    Full Text Available Despite its huge ecological importance, microbial oxygen respiration in pelagic waters is little studied, primarily due to methodological difficulties. Respiration measurements are challenging because of the required high resolution of oxygen concentration measurements. Recent improvements in oxygen sensing techniques bear great potential to overcome these limitations. Here we compare 3 different methods to measure oxygen consumption rates at low oxygen concentrations, utilizing amperometric Clark type sensors (STOX, optical sensors (optodes, and mass spectrometry in combination with (18-18O2 labeling. Oxygen concentrations and consumption rates agreed well between the different methods when applied in the same experimental setting. Oxygen consumption rates between 30 and 400 nmol L(-1 h(-1 were measured with high precision and relative standard errors of less than 3%. Rate detection limits in the range of 1 nmol L(-1 h(-1 were suitable for rate determinations in open ocean water and were lowest at the lowest applied O2 concentration.

  16. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  17. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  18. Formation of neutrophil extracellular traps under low oxygen level

    Directory of Open Access Journals (Sweden)

    Katja Branitzki-Heinemann

    2016-11-01

    Full Text Available Since their discovery, neutrophil extracellular traps (NETs have been characterized as a fundamental host innate immune defense mechanism. Conversely, excessive NET release may have a variety of detrimental consequences for the host. A fine balance between NET formation and elimination is necessary to sustain a protective effect during an infectious challenge. Our own recently published data revealed that stabilization of hypoxia inducible factor 1α (HIF-1α by the iron chelating HIF-1α-agonist desferoxamine or AKB-4924 enhanced the release of phagocyte extracellular traps. Since HIF-1α is a global regulator of the cellular response to low oxygen, we hypothesized that NET formation may be similarly increased under low oxygen conditions. Hypoxia occurs in tissues during infection or inflammation, mostly due to overconsumption of oxygen by pathogens and recruited immune cells. Therefore, experiments were performed to characterize the formation of NETs under hypoxic oxygen conditions compared to normoxia. Human blood-derived neutrophils were isolated and incubated under normoxic (21% oxygen level and compared to hypoxic (1% conditions. Dissolved oxygen levels were monitored in the primary cell culture using a Fibox4-PSt3 measurement system. The formation of NETs was quantified by fluorescence microscopy in response to the known NET-inducer phorbol 12-myristate 13-acetate (PMA or S. aureus wildtype and a nuclease-deficient mutant. In contrast to our hypothesis, spontaneous NET formation of neutrophils incubated under hypoxia was distinctly reduced compared to control neutrophils incubated under normoxia. Furthermore, neutrophils incubated under hypoxia showed significantly reduced formation of NETs in response to PMA. Gene expression analysis revealed that mRNA level of hif-1α as well as hif-1α target genes was not altered. However, in good correlation to the decreased NET formation under hypoxia, the cholesterol content of the neutrophils was

  19. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  20. Hydrogen-rich gas production from waste plastics by pyrolysis and low-temperature steam reforming over a ruthenium catalyst

    International Nuclear Information System (INIS)

    Namioka, Tomoaki; Saito, Atsushi; Inoue, Yukiharu; Park, Yeongsu; Min, Tai-jin; Roh, Seon-ah; Yoshikawa, Kunio

    2011-01-01

    Operating conditions for low-temperature pyrolysis and steam reforming of plastics over a ruthenium catalyst were investigated. In the range studied, the highest gas and lowest coke fractions for polystyrene (PS) with a 60 g h -1 scale, continuous-feed, two-stage gasifier were obtained with a pyrolyzer temperature of 673 K, steam reforming temperature of 903 K, and weight hourly space velocity (WHSV) of 0.10 g-sample g-catalyst -1 h -1 . These operating conditions are consistent with optimum conditions reported previously for polypropylene. Our results indicate that at around 903 K, the activity of the ruthenium catalyst was high enough to minimize the difference between the rates of the steam reforming reactions of the pyrolysates from polystyrene and polypropylene. The proposed system thus has the flexibility to compensate for differences in chemical structures of municipal waste plastics. In addition, the steam reforming temperature was about 200 K lower than the temperature used in a conventional Ni-catalyzed process for the production of hydrogen. Low-temperature steam reforming allows for lower thermal input to the steam reformer, which results in an increase in thermal efficiency in the proposed process employing a Ru catalyst. Because low-temperature steam reforming can be also expected to reduce thermal degradation rates of the catalyst, the pyrolysis-steam reforming process with a Ru catalyst has the potential for use in small-scale production of hydrogen-rich gas from waste plastics that can be used for power generation.

  1. The production of hydrogen-rich gas by wet sludge pyrolysis using waste heat from blast-furnace slag

    International Nuclear Information System (INIS)

    Luo, Siyi; Feng, Yu

    2016-01-01

    Blast furnace (BF) slag, a byproduct of steelmaking industry, contains a large amount of sensible heat and is composed of some metal oxides, which exhibits preferable catalytic performance in improving tar cracking and C_nH_m reforming. This paper presents a heat recovery system from the heat of BF slag, which generates hydrogen-rich gas via the endothermic reactions of sludge pyrolysis. The effects of various parameters including the slag temperature, the mass ratio of slag to sludge (B/S), particle size and feed moisture on product yields and gas characteristics were evaluated separately. It was found that the pyrolysis products distribution was significantly influenced by the BF slag temperature. The differences resulting from varying B/S practically disappear as higher temperature heat carrier is approached. The optimum feed moisture was in favour of sludge pyrolysis by getting char and tar participate in gasification reactions, improving gas yield and quality. BF slag as catalyst can greatly increase H_2 and CO contents of gas by improving tar degradation and reforming of biogas (CO_2 and CH_4). Decreasing the slag particles size was helpful to sludge primary pyrolysis to produce more light gases, less char and condensate, while its effects on gas compositions was not evident. - Highlights: • The sensible heat of molten slag was recovered and converted into combustible gas. • A novel rotary pyrolysis reactor using BF slag as heat carrier was presented. • The moisture in sludge was used as the gasification medium and hydrogen source.

  2. Hydrogen-Rich Syngas Production from Gasification and Pyrolysis of Solar Dried Sewage Sludge: Experimental and Modeling Investigations

    Directory of Open Access Journals (Sweden)

    Aïda Ben Hassen Trabelsi

    2017-01-01

    Full Text Available Solar dried sewage sludge (SS conversion by pyrolysis and gasification processes has been performed, separately, using two laboratory-scale reactors, a fixed-bed pyrolyzer and a downdraft gasifier, to produce mainly hydrogen-rich syngas. Prior to SS conversion, solar drying has been conducted in order to reduce moisture content (up to 10%. SS characterization reveals that these biosolids could be appropriate materials for gaseous products production. The released gases from SS pyrolysis and gasification present relatively high heating values (up to 9.96 MJ/kg for pyrolysis and 8.02  9.96 MJ/kg for gasification due to their high contents of H2 (up to 11 and 7 wt%, resp. and CH4 (up to 17 and 5 wt%, resp.. The yields of combustible gases (H2 and CH4 show further increase with pyrolysis. Stoichiometric models of both pyrolysis and gasification reactions were determined based on the global biomass formula, CαHβOγNδSε, in order to assist in the products yields optimization.

  3. Performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas

    Energy Technology Data Exchange (ETDEWEB)

    Roy, M.M.; Tomita, E.; Kawahara, N.; Harada, Y.; Sakane, A. [Okayama University, Okayama (Japan). Dept. of Mechanical Engineering

    2009-12-15

    This study investigated the engine performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas and ignited by a pilot amount of diesel fuel. The engine was tested for use as a cogeneration engine, so power output while maintaining a reasonable thermal efficiency was important. Experiments were carried out at a constant pilot injection pressure and pilot quantity for different fuel-air equivalence ratios and at various injection timings without and with exhaust gas recirculation (EGR). The experimental strategy was to optimize the injection timing to maximize engine power at different fuel-air equivalence ratios without knocking and within the limit of the maximum cylinder pressure. The engine was tested first without EGR condition up to the maximum possible fuel-air equivalence ratio of 0.65. A maximum indicated mean effective pressure (IMEP) of 1425 kPa and a thermal efficiency of 39% were obtained. However, the nitrogen oxides (NOx) emissions were high. A simulated EGR up to 50% was then performed to obtain lower NOx emissions. The maximum reduction of NOx was 60% or more maintaining the similar levels of IMEP and thermal efficiency. Two-stage combustion was obtained; this is an indicator of maximum power output conditions and a precursor of knocking combustion.

  4. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  5. Cell wide responses to low oxygen exposure in Desulfovibriovulgaris Hildenborough

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, A.; Redding, A.; Joachimiak, M.; Arkin, A.; Borglin, S.; Dehal, P.; Chakraborty, R.; Geller, J.; Hazen, T.; He, Q.; Joyner, D.; Martin, V.; Wall, J.; Yang, Z.; Zhou, J.; Keasling, J.

    2007-03-11

    The responses of the anaerobic, sulfate-reducing Desulfovibrio vulgaris Hildenborough to low oxygen exposure (0.1% O{sub 2}) were monitored via transcriptomics and proteomics. Exposure to 0.1% O{sub 2} caused a decrease in growth rate without affecting viability. A concerted up regulation in the predicted peroxide stress response regulon (PerR) genes was observed in response to the 0.1% O{sub 2} exposure. Several of these candidates also showed increases in protein abundance. Among the remaining small number of transcript changes was the up regulation of the predicted transmembrane tetraheme cytochrome c3 complex. Other known oxidative stress response candidates remained unchanged during this low O{sub 2} exposure. To fully understand the results of the 0.1% O{sub 2} exposure, transcriptomics and proteomics data were collected for exposure to air using a similar experimental protocol. In contrast to the 0.1% O{sub 2} exposure, air exposure was detrimental to both the growth rate and viability and caused dramatic changes at both the transcriptome and proteome levels. Interestingly, the transcripts of the predicted PerR regulon genes were down regulated during air exposure. Our results highlight the differences in the cell wide response to low and high O{sub 2} levels of in D. vulgaris and suggest that while exposure to air is highly detrimental to D. vulgaris, this bacterium can successfully cope with periodic exposure to low O{sub 2} levels in its environment.

  6. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  7. Co-pyrolysis of waste tire/coal mixtures for smokeless fuel, maltenes and hydrogen-rich gas production

    International Nuclear Information System (INIS)

    Bičáková, Olga; Straka, Pavel

    2016-01-01

    Highlights: • Co-pyrolysis of waste tires/coal mixtures yields mainly smokeless fuel (55–74 wt%). • Alternatively, the smokeless fuel can serve as carbonaceous sorbent. • The obtained tar contained maltenes (80–85 wt%) and asphaltenes (6–8 wt%). • Tar from co-pyrolysis can serve as heating oil or a source of maltenes for repairing of asphalt surfaces. • The hydrogen-rich gas was obtained (61–65 vol% H_2, 24–25 vol% CH_4, 1.4–2 vol% CO_2). - Abstract: The processing of waste tires with two different types of bituminous coal was studied through the slow co-pyrolysis of 1 kg of waste tire/coal mixtures with 15, 30 and 60 wt% waste tires on a laboratory scale. The waste tire/coal mixtures were pyrolysed using a quartz reactor in a stationary bed. The mixtures were heated at a rate 5 °C/min up to the final temperature of 900 °C with a soaking time of 30 min at the required temperature. The mass balance of the process and the properties of the coke and tar obtained were evaluated, further, the influence of the admixture in the charge on the amount and composition of the obtained coke and tar was determined. It was found that the smokeless fuel/carbonaceous sorbent and a high yield of tar for further use can be obtained through the slow co-pyrolysis. The obtained tars contained mostly maltenes (80–85 wt%). FTIR analysis showed that the maltenes from the co-pyrolysis of coal/waste tires exhibited significantly lower aromaticity as compared with that from coal alone. The gas obtained from pyrolysis or co-pyrolysis of waste tire/coal mixtures contained a high amount of hydrogen (above 60 vol%) and methane (above 20 vol%).

  8. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  9. Hydrogen-rich saline controls remifentanil-induced hypernociception and NMDA receptor NR1 subunit membrane trafficking through GSK-3β in the DRG in rats.

    Science.gov (United States)

    Zhang, Linlin; Shu, Ruichen; Wang, Chunyan; Wang, Haiyun; Li, Nan; Wang, Guolin

    2014-07-01

    Although NMDAR trafficking mediated by GSK-3β involvement in transmission of pronociceptive messages in the spinal cord has been confirmed by our previous studies, whether NMDAR trafficking is implicated in peripheral sensitization remains equivocal. It is demonstrated that inflammation is associated with spinal NMDAR-containing nociceptive neurons activation and the maintenance of opioid induced pain hypersensitivity. However, whether and how hydrogen-rich saline, as an effective anti-inflammatory drug, could prevent hyperalgesia through affecting peripheral sensitization caused by NMDAR activation remains to be explored. To test these effects, hydrogen-rich saline (2.5, 5 or 10 ml/kg) was administrated intraperitoneally after remifentanil infusion, NMDAR antagonist MK-801 or GSK-3β inhibitor TDZD-8 was administrated intravenously before remifentanil infusion in rats. We examined time course of hydrogen concentration in blood after hydrogen-rich saline administration. Mechanical and thermal hyperalgesia were evaluated by measuring PWT and PWL for 48 post-infusion hours, respectively. Western blotting and real-time qPCR assay were applied to analyze the NR1 membrane trafficking, GSK-3β expression and activity in DRG. Inflammatory mediators (TNF-α, IL-1β, and IL-6) expressions in DRG were also analyzed. We found that NR1 membrane trafficking in DRG increased, possibly due to GSK-3β activation after remifentanil infusion. We also discovered that hydrogen-rich saline not 2.5 ml/kg but 5 and 10 ml/kg could dose-dependently attenuate mechanical and thermal hyperalgesia without affecting baseline nociceptive threshold, reduce expressions of inflammatory mediators (TNF-α, IL-1β, and IL-6) and decrease NR1 trafficking mediated by GSK-3β, and minimal effective concentration was observed to be higher than 10 μmol/L, namely peak concentration in arterial blood after administration of HRS 2.5 ml/kg without any influence on hyperalgesia. Our results indicated that

  10. Development of a Low NOx Medium sized Industrial Gas Turbine Operating on Hydrogen-Rich Renewable and Opportunity Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Srinivasan, Ram

    2013-07-31

    This report presents the accomplishments at the completion of the DOE sponsored project (Contract # DE-FC26-09NT05873) undertaken by Solar Turbines Incorporated. The objective of this 54-month project was to develop a low NOx combustion system for a medium sized industrial gas turbine engine operating on Hydrogen-rich renewable and opportunity Fuels. The work in this project was focused on development of a combustion system sized for 15MW Titan 130 gas turbine engine based on design analysis and rig test results. Although detailed engine evaluation of the complete system is required prior to commercial application, those tasks were beyond the scope of this DOE sponsored project. The project tasks were organized in three stages, Stages 2 through 4. In Stage 2 of this project, Solar Turbines Incorporated characterized the low emission capability of current Titan 130 SoLoNOx fuel injector while operating on a matrix of fuel blends with varying Hydrogen concentration. The mapping in this phase was performed on a fuel injector designed for natural gas operation. Favorable test results were obtained in this phase on emissions and operability. However, the resulting fuel supply pressure needed to operate the engine with the lower Wobbe Index opportunity fuels would require additional gas compression, resulting in parasitic load and reduced thermal efficiency. In Stage 3, Solar characterized the pressure loss in the fuel injector and developed modifications to the fuel injection system through detailed network analysis. In this modification, only the fuel delivery flowpath was modified and the air-side of the injector and the premixing passages were not altered. The modified injector was fabricated and tested and verified to produce similar operability and emissions as the Stage 2 results. In parallel, Solar also fabricated a dual fuel capable injector with the same air-side flowpath to improve commercialization potential. This injector was also test verified to produce 15

  11. [The protection of hydrogen-rich saline on a rat dry eye model induced by scopolamine hydrobromide].

    Science.gov (United States)

    Chu, Y Y; Hua, N; Ru, Y S; Zhao, S Z

    2017-05-11

    Objective: To evaluate the effect of hydrogen-rich saline (HRS) on dry eye rats induced by subcutaneous injection of scopolamine hydrobromide. Methods: Experiment research. Thirty female Wistar rats at about six weeks old were randomly divided into the normal group, dry eye group, HRS eyedrops group, normal saline eyedrops group (NS), HRS intraperitoneal injection group and NS intraperitoneal injection group, with 5 rats in each group. The dry eye was induced by subcutaneous injection of scopolamine hydrobromide in the latter five groups. The clinical signs of dry eye such as tear volume (SⅠt), tear break-up time (BUT) and corneal epithelial fluorescein staining scores were evaluated on day 7, 14, 21 and 28. On the 28th day, ten eyes in each group were enucleated and processed for paraffin sections for HE, PAS and immunohistochemistry stainings. Analysis of variance was used to test the data, and independent samples t -test was used for comparison between the two groups. Two-way repeated measure ANOVA was used to compare the difference among groups at different time points, one-way ANOVA was used to test the comparisons of the clinical signs at one time, and LSD was used to for comparison between two groups. Results: Before and after the experiment of the day 7, 14, 21, 28, the values of SIt in HRS eyedrops group and HRS intraperitoneal injection group were respectively:(3.625±1.157),(3.313±0.704),(3.250±0.535),(3.313±0.372), (3.375±0.582)mm and (3.500±1.019), (2.893±0.656), (3.321±0.668), (3.179±0.575), (3.214±0.871)mm. The values of BUT were respectively: (2.750±0.707), (2.688±0.594), (2.813±0.753), (3.000±0.756), (2.750±0.707)s and (3.000±0.679), (2.321±0.464), (2.750±0.753), (3.214±0.699), (2.679±0.608)s. The values of fluorescein staining score were respectively: (6.250±0.707), (8.875±0.641), (8.750±0.707), (9.250±0.463), (8.250±1.282) and (6.000±0.679), (9.143±1.027), (8.857±0.770), (9.143±0.949), (8.500±0.760). The difference

  12. Hydrogen-rich Water Exerting a Protective Effect on Ovarian Reserve Function in a Mouse Model of Immune Premature Ovarian Failure Induced by Zona Pellucida 3

    Science.gov (United States)

    He, Xin; Wang, Shu-Yu; Yin, Cheng-Hong; Wang, Tong; Jia, Chan-Wei; Ma, Yan-Min

    2016-01-01

    Background: Premature ovarian failure (POF) is a disease that affects female fertility but has few effective treatments. Ovarian reserve function plays an important role in female fertility. Recent studies have reported that hydrogen can protect male fertility. Therefore, we explored the potential protective effect of hydrogen-rich water on ovarian reserve function through a mouse immune POF model. Methods: To set up immune POF model, fifty female BALB/c mice were randomly divided into four groups: Control (mice consumed normal water, n = 10), hydrogen (mice consumed hydrogen-rich water, n = 10), model (mice were immunized with zona pellucida glycoprotein 3 [ZP3] and consumed normal water, n = 15), and model-hydrogen (mice were immunized with ZP3 and consumed hydrogen-rich water, n = 15) groups. After 5 weeks, mice were sacrificed. Serum anti-Müllerian hormone (AMH) levels, granulosa cell (GC) apoptotic index (AI), B-cell leukemia/lymphoma 2 (Bcl-2), and BCL2-associated X protein (Bax) expression were examined. Analyses were performed using SPSS 17.0 (SPSS Inc., Chicago, IL, USA) software. Results: Immune POF model, model group exhibited markedly reduced serum AMH levels compared with those of the control group (5.41 ± 0.91 ng/ml vs. 16.23 ± 1.97 ng/ml, P = 0.033) and the hydrogen group (19.65 ± 7.82 ng/ml, P = 0.006). The model-hydrogen group displayed significantly higher AMH concentrations compared with that of the model group (15.03 ± 2.75 ng/ml vs. 5.41 ± 0.91 ng/ml, P = 0.021). The GC AI was significantly higher in the model group (21.30 ± 1.74%) than those in the control (7.06 ± 0.27%), hydrogen (5.17 ± 0.41%), and model-hydrogen groups (11.24 ± 0.58%) (all P hydrogen group compared with that of the hydrogen group (11.24 ± 0.58% vs. 5.17 ± 0.41%, P = 0.021). Compared with those of the model group, ovarian tissue Bcl-2 levels increased (2.18 ± 0.30 vs. 3.01 ± 0.33, P = 0.045) and the Bax/Bcl-2 ratio decreased in the model-hydrogen group

  13. Low oxygen level increases proliferation and metabolic changes in bovine granulosa cells.

    Science.gov (United States)

    Shiratsuki, Shogo; Hara, Tomotaka; Munakata, Yasuhisa; Shirasuna, Koumei; Kuwayama, Takehito; Iwata, Hisataka

    2016-12-05

    The present study addresses molecular backgrounds underlying low oxygen induced metabolic changes and 1.2-fold change in bovine granulosa cell (GCs) proliferation. RNA-seq revealed that low oxygen (5%) upregulated genes associated with HIF-1 and glycolysis and downregulated genes associated with mitochondrial respiration than that in high oxygen level (21%). Low oxygen level induced high glycolytic activity and low mitochondrial function and biogenesis. Low oxygen level enhanced GC proliferation with high expression levels of HIF-1, VEGF, AKT, mTOR, and S6RP, whereas addition of anti-VEGF antibody decreased cellular proliferation with low phosphorylated AKT and mTOR expression levels. Low oxygen level reduced SIRT1, whereas activation of SIRT1 by resveratrol increased mitochondrial replication and decreased cellular proliferation with reduction of phosphorylated mTOR. These results suggest that low oxygen level stimulates the HIF1-VEGF-AKT-mTOR pathway and up-regulates glycolysis, which contributes to GC proliferation, and downregulation of SIRT1 contributes to hypoxia-associated reduction of mitochondria and cellular proliferation. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

  14. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  15. Hydrogen-rich medium protects mouse embryonic fibroblasts from oxidative stress by activating LKB1-AMPK-FoxO1 signal pathway.

    Science.gov (United States)

    Lee, Jihyun; Yang, Goowon; Kim, Young-Joo; Tran, Quynh Hoa; Choe, Wonchae; Kang, Insug; Kim, Sung Soo; Ha, Joohun

    2017-09-23

    Persistent oxidative stress is recognized as a major cause of many pathological conditions as well as ageing. However, most clinical trials of dietary antioxidants have failed to produce successful outcomes in treating oxidative stress-induced diseases. Molecular hydrogen (H 2 ) has recently received considerable attention as a therapeutic agent owing to its novel antioxidant properties, a selective scavenger of hydroxyl and peroxynitrite radicals. Beyond this, numerous reports support that H 2 can modulate the activity of various cellular signal pathways. However, its effect on AMP-activated protein kinase (AMPK) signal pathway, a central regulator of energy hemostasis, has remained almost elusive. Here, we report that hydrogen-rich medium activated LKB1-AMPK signal pathway without ATP depletion, which in turn induced FoxO1-dependent transcription of manganese superoxide dismutase and catalase in mouse embryonic fibroblasts. Moreover, hydrogen-rich media effectively reduced the level of reactive oxygen species in cells treated with hydrogen peroxide and protected these cells from apoptosis in an AMPK-dependent manner. These results suggest that the LKB1-AMPK-FoxO1 signaling pathway is a critical mediator of the antioxidant properties of H 2 , further supporting the idea that H 2 acts as a signaling molecule to serve various physiological functions. Copyright © 2017 Elsevier Inc. All rights reserved.

  16. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  17. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

  18. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

    Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

  19. Low oxygen biomass-derived pyrolysis oils and methods for producing the same

    Science.gov (United States)

    Marinangeli, Richard; Brandvold, Timothy A; Kocal, Joseph A

    2013-08-27

    Low oxygen biomass-derived pyrolysis oils and methods for producing them from carbonaceous biomass feedstock are provided. The carbonaceous biomass feedstock is pyrolyzed in the presence of a catalyst comprising base metal-based catalysts, noble metal-based catalysts, treated zeolitic catalysts, or combinations thereof to produce pyrolysis gases. During pyrolysis, the catalyst catalyzes a deoxygenation reaction whereby at least a portion of the oxygenated hydrocarbons in the pyrolysis gases are converted into hydrocarbons. The oxygen is removed as carbon oxides and water. A condensable portion (the vapors) of the pyrolysis gases is condensed to low oxygen biomass-derived pyrolysis oil.

  20. Catalytic steam gasification of biomass in fluidized bed at low temperature: Conversion from livestock manure compost to hydrogen-rich syngas

    International Nuclear Information System (INIS)

    Xiao, Xianbin; Le, Duc Dung; Li, Liuyun; Meng, Xianliang; Cao, Jingpei; Morishita, Kayoko; Takarada, Takayuki

    2010-01-01

    Utilizing large amounts of animal waste as a source of renewable energy has the potential to reduce its disposal problems and associated pollution issues. Gasification characteristics of the manure compost make it possible for low temperature gasification. In this paper, an energy efficient approach to hydrogen-rich syngas from manure compost is represented at relatively low temperature, around 600 o C, in a continuous-feeding fluidized bed reactor. The effects of catalyst performance, reactor temperature, steam, and reaction type on gas yield, gas composition, and carbon conversion efficiency are discussed. The Ni-Al 2 O 3 catalyst simultaneously promotes tar cracking and steam reforming. Higher temperature contributes to higher gas yield and carbon conversion. The steam introduction increases hydrogen yield, by steam reforming and water-gas shift reaction. Two-stage gasification is also tried, showing the advantage of better catalyst utilization and enhancing the catalytic reactions to some extent.

  1. Co-pyrolysis of coal with hydrogen-rich gases. 1. Coal pyrolysis under coke-oven gas and synthesis gas

    Energy Technology Data Exchange (ETDEWEB)

    Liao, H.; Li, B.; Zhang, B. [Chinese Academy of Sciences, Taiyuan (China). State Key Lab. of Coal Conversion

    1998-06-01

    To improve the economics of the hydropyrolysis process, it has been suggested that cheaper hydrogen-rich gases (such as coke oven gas, synthesis gas) could be used instead of pure hydrogen. Pyrolysis of Chinese Xianfeng lignite was carried out with coke oven gas (COG) and synthesis gas (SG) as reactive gases at 0.1-5 MPa and at a final temperature up to 650{degree}C with a heating rate of 5-25{degree}C min{sup -1} in a 10 g fixed-bed reactor. The results indicate that it is possible to use COG and SG instead of pure hydrogen in hydropyrolysis, but that the experimental conditions must be adjusted to optimize the yields of the valuable chemicals. 14 refs., 3 figs., 6 tabs.

  2. Accelerated ceria–zirconia solubilization by cationic diffusion inversion at low oxygen activity

    DEFF Research Database (Denmark)

    Esposito, Vincenzo; Ni, De Wei; Marani, Debora

    2016-01-01

    Fast elemental diffusion at the Gd-doped ceria/Y-stabilized zirconia interface occurs under reducing conditions at low oxygen activity (pO2 < 10−12 atm) and high temperature (1400 °C). This effect leads to formation of thick ceria–zirconia solid solution reaction layers in the micro-range vs. thi...

  3. Characterization of in vitro chlamydial cultures in low-oxygen atmospheres

    DEFF Research Database (Denmark)

    Juul, Nicolai Stefan; Jensen, Helene; Hvid, Malene

    2007-01-01

    To mimic in vivo conditions during chlamydial infections, Chlamydia trachomatis serovar D and Chlamydia pneumoniae CWL029 were cultured in low-oxygen atmospheres containing 4% O(2), with parallel controls cultured in atmospheric air. Both were enriched with 5% CO(2). The results showed a dramatic...

  4. Macrophages Under Low Oxygen Culture Conditions Respond to Ion Parametric Resonance Magnetic Fields

    Science.gov (United States)

    Macrophages, when entering inflamed tissue, encounter low oxygen tension due to the impairment of blood supply and/or the massive infiltration of cells that consume oxygen. Previously, we showed that such macrophages release more bacteriotoxic hydrogen peroxide (H202) when expose...

  5. The Effect of Low Oxygen Stress on Phytophthora cinnamomi Infection and Disease of Cork Oak Roots

    Science.gov (United States)

    Karel A. Jacobs; James D. MacDonald; Alison M. Berry; Laurence R. Costello

    1997-01-01

    The incidence and severity of Phytophthora cinnamomi Rands root disease was quantified in cork oak (Quercus suber L.) roots subjected to low oxygen (hypoxia) stress. Seedling root tips were inoculated with mycelial plugs of the fungus and incubated in ≤1, 3-4, or 21 percent oxygen for 5 days. Ninety-four percent of roots...

  6. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

    This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

  7. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  8. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corractive measures taken are also described

  9. Detection of ultra-low oxygen concentration based on the fluorescence blinking dynamics of single molecules

    Science.gov (United States)

    Wu, Ruixiang; Chen, Ruiyun; Zhou, Haitao; Qin, Yaqiang; Zhang, Guofeng; Qin, Chengbing; Gao, Yan; Gao, Yajun; Xiao, Liantuan; Jia, Suotang

    2018-01-01

    We present a sensitive method for detection of ultra-low oxygen concentrations based on the fluorescence blinking dynamics of single molecules. The relationship between the oxygen concentration and the fraction of time spent in the off-state, stemming from the population and depopulation of triplet states and radical cationic states, can be fitted with a two-site quenching model in the Stern-Volmer plot. The oxygen sensitivity is up to 43.42 kPa-1 in the oxygen partial pressure region as low as 0.01-0.25 kPa, which is seven times higher than that of the fluorescence intensity indicator. This method avoids the limitation of the sharp and non-ignorable fluctuations that occur during the measurement of fluorescence intensity, providing potential applications in the field of low oxygen-concentration monitoring in life science and industry.

  10. Effect of low oxygen tension on the biological characteristics of human bone marrow mesenchymal stem cells

    OpenAIRE

    Kim, Dae Seong; Ko, Young Jong; Lee, Myoung Woo; Park, Hyun Jin; Park, Yoo Jin; Kim, Dong-Ik; Sung, Ki Woong; Koo, Hong Hoe; Yoo, Keon Hee

    2016-01-01

    Culture of mesenchymal stem cells (MSCs) under ambient conditions does not replicate the low oxygen environment of normal physiological or pathological states and can result in cellular impairment during culture. To overcome these limitations, we explored the effect of hypoxia (1 % O2) on the biological characteristics of MSCs over the course of different culture periods. The following biological characteristics were examined in human bone marrow-derived MSCs cultured under hypoxia for 8 week...

  11. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  12. Two-step gasification of cattle manure for hydrogen-rich gas production: Effect of biochar preparation temperature and gasification temperature.

    Science.gov (United States)

    Xin, Ya; Cao, Hongliang; Yuan, Qiaoxia; Wang, Dianlong

    2017-10-01

    Two-step gasification process was proposed to dispose cattle manure for hydrogen rich gas production. The effect of temperature on product distribution and biochar properties were first studied in the pyrolysis-carbonization process. The steam gasification of biochar derived from different pyrolysis-carbonization temperatures was then performed at 750°C and 850°C. The biochar from the pyrolysis-carbonization temperatures of 500°C had high carbon content and low volatiles content. According to the results of gasification stage, the pyrolysis-carbonization temperature of 500°C and the gasification temperature of 850°C were identified as the suitable conditions for hydrogen production. We obtained 1.61m 3 /kg of syngas production, 0.93m 3 /kg of hydrogen yield and 57.58% of hydrogen concentration. This study shows that two-step gasification is an efficient waste-to-hydrogen energy process. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Promotion of hydrogen-rich gas and phenolic-rich bio-oil production from green macroalgae Cladophora glomerata via pyrolysis over its bio-char.

    Science.gov (United States)

    Norouzi, Omid; Jafarian, Sajedeh; Safari, Farid; Tavasoli, Ahmad; Nejati, Behnam

    2016-11-01

    Conversion of Cladophora glomerata (C. glomerata) as a Caspian Sea's green macroalgae into gaseous, liquid and solid products was carried out via pyrolysis at different temperatures to determine its potential for bio-oil and hydrogen-rich gas production for further industrial utilization. Non-catalytic tests were performed to determine the optimum condition for bio-oil production. The highest portion of bio-oil was retrieved at 500°C. The catalytic test was performed using the bio-char derived at 500°C as a catalyst. Effect of the addition of the algal bio-char on the composition of the bio-oil and also gaseous products was investigated. Pyrolysis derived bio-char was characterized by BET, FESEM and ICP method to show its surface area, porosity, and presence of inorganic metals on its surface, respectively. Phenols were increased from 8.5 to 20.76area% by the addition of bio-char. Moreover, the hydrogen concentration and hydrogen selectivity were also enhanced by the factors of 1.37, 1.59 respectively. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  15. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  16. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  17. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  18. Effect of reduced light and low oxygen concentration on germination, growth and establishment of some plants

    DEFF Research Database (Denmark)

    Yasin, Muhammad

    Many abiotic factors effect plants germination, growth, and development. This Ph.D. study elucidates the effect of reduced light, low oxygen and seed dormancy on germination and growth of some weed species, field crops and vegetables. One study describes the growth and developmental responses...... of some common, invasive and rare weed species to reduced light levels in greenhouse experiments. The seed germination response of some weed species, field crops, and vegetables to different oxygen concentrations was also quantified in the laboratory experiments. The effect of east-west (EW) and north...

  19. Anti-oxidant and anti-inflammatory effects of hydrogen-rich water alleviate ethanol-induced fatty liver in mice.

    Science.gov (United States)

    Lin, Ching-Pin; Chuang, Wen-Chen; Lu, Fung-Jou; Chen, Chih-Yen

    2017-07-21

    To investigate the effects of hydrogen-rich water (HRW) treatment on prevention of ethanol (EtOH)-induced early fatty liver in mice. In vitro reduction of hydrogen peroxide by HRW was determined with a chemiluminescence system. Female mice were randomly divided into five groups: control, EtOH, EtOH + silymarin, EtOH + HRW and EtOH + silymarin + HRW. Each group was fed a Lieber-DeCarli liquid diet containing EtOH or isocaloric maltose dextrin (control diet). Silymarin was used as a positive control to compare HRW efficacy against chronic EtOH-induced hepatotoxicity. HRW was freshly prepared and given at a dosage of 1.2 mL/mouse trice daily. Blood and liver tissue were collected after chronic-binge liquid-diet feeding for 12 wk. The in vitro study showed that HRW directly scavenged hydrogen peroxide. The in vivo study showed that HRW increased expression of acyl ghrelin, which was correlated with food intake. HRW treatment significantly reduced EtOH-induced increases in serum alanine aminotransferase, aspartate aminotransferase, triglycerol and total cholesterol levels, hepatic lipid accumulation and inflammatory cytokines, including tumor necrosis factor-alpha (TNF-α) and interleukin (IL)-6. HRW attenuated malondialdehyde level, restored glutathione depletion and increased superoxide dismutase, glutathione peroxidase and catalase activities in the liver. Moreover, HRW reduced TNF-α and IL-6 levels but increased IL-10 and IL-22 levels. HRW protects against chronic EtOH-induced liver injury, possibly by inducing acyl ghrelin to suppress the pro-inflammatory cytokines TNF-α and IL-6 and induce IL-10 and IL-22, thus activating antioxidant enzymes against oxidative stress.

  20. XUV-exposed, non-hydrostatic hydrogen-rich upper atmospheres of terrestrial planets. Part I: atmospheric expansion and thermal escape.

    Science.gov (United States)

    Erkaev, Nikolai V; Lammer, Helmut; Odert, Petra; Kulikov, Yuri N; Kislyakova, Kristina G; Khodachenko, Maxim L; Güdel, Manuel; Hanslmeier, Arnold; Biernat, Helfried

    2013-11-01

    The recently discovered low-density "super-Earths" Kepler-11b, Kepler-11f, Kepler-11d, Kepler-11e, and planets such as GJ 1214b represent the most likely known planets that are surrounded by dense H/He envelopes or contain deep H₂O oceans also surrounded by dense hydrogen envelopes. Although these super-Earths are orbiting relatively close to their host stars, they have not lost their captured nebula-based hydrogen-rich or degassed volatile-rich steam protoatmospheres. Thus, it is interesting to estimate the maximum possible amount of atmospheric hydrogen loss from a terrestrial planet orbiting within the habitable zone of late main sequence host stars. For studying the thermosphere structure and escape, we apply a 1-D hydrodynamic upper atmosphere model that solves the equations of mass, momentum, and energy conservation for a planet with the mass and size of Earth and for a super-Earth with a size of 2 R(Earth) and a mass of 10 M(Earth). We calculate volume heating rates by the stellar soft X-ray and extreme ultraviolet radiation (XUV) and expansion of the upper atmosphere, its temperature, density, and velocity structure and related thermal escape rates during the planet's lifetime. Moreover, we investigate under which conditions both planets enter the blow-off escape regime and may therefore experience loss rates that are close to the energy-limited escape. Finally, we discuss the results in the context of atmospheric evolution and implications for habitability of terrestrial planets in general.

  1. Hydrogen-rich water inhibits glucose and α,β -dicarbonyl compound-induced reactive oxygen species production in the SHR.Cg-Leprcp/NDmcr rat kidney

    Directory of Open Access Journals (Sweden)

    Katakura Masanori

    2012-07-01

    Full Text Available Abstract Background Reactive oxygen species (ROS production induced by α,β-dicarbonyl compounds and advanced glycation end products causes renal dysfunction in patients with type 2 diabetes and metabolic syndrome. Hydrogen-rich water (HRW increases the H2 level in blood and tissues, thus reducing oxidative stress in animals as well as humans. In this study, we investigated the effects of HRW on glucose- and α,β-dicarbonyl compound-induced ROS generation in vitro and in vivo. Methods Kidney homogenates from Wistar rats were incubated in vitro with glucose and α,β-dicarbonyl compounds containing HRW, following which ROS levels were measured. In vivo animal models of metabolic syndrome, SHR.Cg-Leprcp/NDmcr rats, were treated with HRW for 16 weeks, following which renal ROS production and plasma and renal α,β-dicarbonyl compound levels were measured by liquid chromatograph mass spectrometer. Results HRW inhibited glucose- and α,β-dicarbonyl compound-induced ROS production in kidney homogenates from Wistar rats in vitro. Furthermore, SHR.Cg-Leprcp/NDmcr rats treated with HRW showed a 34% decrease in ROS production. Moreover, their renal glyoxal, methylglyoxal, and 3-deoxyglucosone levels decreased by 81%, 77%, and 60%, respectively. Positive correlations were found between renal ROS levels and renal glyoxal (r = 0.659, p = 0.008 and methylglyoxal (r = 0.782, p = 0.001 levels. Conclusion These results indicate that HRW inhibits the production of α,β-dicarbonyl compounds and ROS in the kidneys of SHR.Cg-Leprcp/NDmcr rats. Therefore, it has therapeutic potential for renal dysfunction in patient with type 2 diabetes and metabolic syndrome.

  2. Effect of low oxygen tension on the biological characteristics of human bone marrow mesenchymal stem cells.

    Science.gov (United States)

    Kim, Dae Seong; Ko, Young Jong; Lee, Myoung Woo; Park, Hyun Jin; Park, Yoo Jin; Kim, Dong-Ik; Sung, Ki Woong; Koo, Hong Hoe; Yoo, Keon Hee

    2016-11-01

    Culture of mesenchymal stem cells (MSCs) under ambient conditions does not replicate the low oxygen environment of normal physiological or pathological states and can result in cellular impairment during culture. To overcome these limitations, we explored the effect of hypoxia (1 % O 2 ) on the biological characteristics of MSCs over the course of different culture periods. The following biological characteristics were examined in human bone marrow-derived MSCs cultured under hypoxia for 8 weeks: proliferation rate, morphology, cell size, senescence, immunophenotypic characteristics, and the expression levels of stemness-associated factors and cytokine and chemokine genes. MSCs cultured under hypoxia for approximately 2 weeks showed increased proliferation and viability. During long-term culture, hypoxia delayed phenotypic changes in MSCs, such as increased cell volume, altered morphology, and the expression of senescence-associated-β-gal, without altering their characteristic immunophenotypic characteristics. Furthermore, hypoxia increased the expression of stemness and chemokine-related genes, including OCT4 and CXCR7, and did not decrease the expression of KLF4, C-MYC, CCL2, CXCL9, CXCL10, and CXCR4 compared with levels in cells cultured under normoxia. In conclusion, low oxygen tension improved the biological characteristics of MSCs during ex vivo expansion. These data suggest that hypoxic culture could be a useful method for increasing the efficacy of MSC cell therapies.

  3. High or low oxygen saturation and severe retinopathy of prematurity: a meta-analysis.

    Science.gov (United States)

    Chen, Minghua L; Guo, Lei; Smith, Lois E H; Dammann, Christiane E L; Dammann, Olaf

    2010-06-01

    Low oxygen saturation appears to decrease the risk of severe retinopathy of prematurity (ROP) in preterm newborns when administered during the first few weeks after birth. High oxygen saturation seems to reduce the risk at later postmenstrual ages (PMAs). However, previous clinical studies are not conclusive individually. To perform a systematic review and meta-analysis to report the association between severe ROP incidence of premature infants with high or low target oxygen saturation measured by pulse oximetry. Studies were identified through PubMed and Embase literature searches through May 2009 by using the terms "retinopathy of prematurity and oxygen" or "retinopathy of prematurity and oxygen therapy." We selected 10 publications addressing the association between severe ROP and target oxygen saturation measured by pulse oximetry. Using a random-effects model we calculated the summary-effect estimate. We visually inspected funnel plots to examine possible publication bias. Low oxygen saturation (70%-96%) in the first several postnatal weeks was associated with a reduced risk of severe ROP (risk ratio [RR]: 0.48 [95% confidence interval (CI): 0.31-0.75]). High oxygen saturation (94%-99%) at > or = 32 weeks' PMA was associated with a decreased risk for progression to severe ROP (RR: 0.54 [95% CI: 0.35-0.82]). Among preterm infants with a gestational age of large randomized clinical trial with long-term developmental follow-up is warranted to confirm this meta-analytic result.

  4. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  5. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  6. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  7. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  8. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  9. XUV-Exposed, Non-Hydrostatic Hydrogen-Rich Upper Atmospheres of Terrestrial Planets. Part II: Hydrogen Coronae and Ion Escape

    Science.gov (United States)

    Lammer, Helmut; Holmström, Mats; Panchenko, Mykhaylo; Odert, Petra; Erkaev, Nikolai V.; Leitzinger, Martin; Khodachenko, Maxim L.; Kulikov, Yuri N.; Güdel, Manuel; Hanslmeier, Arnold

    2013-01-01

    Abstract We studied the interactions between the stellar wind plasma flow of a typical M star, such as GJ 436, and the hydrogen-rich upper atmosphere of an Earth-like planet and a “super-Earth” with a radius of 2 REarth and a mass of 10 MEarth, located within the habitable zone at ∼0.24 AU. We investigated the formation of extended atomic hydrogen coronae under the influences of the stellar XUV flux (soft X-rays and EUV), stellar wind density and velocity, shape of a planetary obstacle (e.g., magnetosphere, ionopause), and the loss of planetary pickup ions on the evolution of hydrogen-dominated upper atmospheres. Stellar XUV fluxes that are 1, 10, 50, and 100 times higher compared to that of the present-day Sun were considered, and the formation of high-energy neutral hydrogen clouds around the planets due to the charge-exchange reaction under various stellar conditions was modeled. Charge-exchange between stellar wind protons with planetary hydrogen atoms, and photoionization, lead to the production of initially cold ions of planetary origin. We found that the ion production rates for the studied planets can vary over a wide range, from ∼1.0×1025 s−1 to ∼5.3×1030 s−1, depending on the stellar wind conditions and the assumed XUV exposure of the upper atmosphere. Our findings indicate that most likely the majority of these planetary ions are picked up by the stellar wind and lost from the planet. Finally, we estimated the long-time nonthermal ion pickup escape for the studied planets and compared them with the thermal escape. According to our estimates, nonthermal escape of picked-up ionized hydrogen atoms over a planet's lifetime within the habitable zone of an M dwarf varies between ∼0.4 Earth ocean equivalent amounts of hydrogen (EOH) to stars—Early atmospheres—Earth-like exoplanets—Energetic neutral atoms—Ion escape—Habitability. Astrobiology 13, 1030–1048. PMID:24283926

  10. XUV-exposed, non-hydrostatic hydrogen-rich upper atmospheres of terrestrial planets. Part II: hydrogen coronae and ion escape.

    Science.gov (United States)

    Kislyakova, Kristina G; Lammer, Helmut; Holmström, Mats; Panchenko, Mykhaylo; Odert, Petra; Erkaev, Nikolai V; Leitzinger, Martin; Khodachenko, Maxim L; Kulikov, Yuri N; Güdel, Manuel; Hanslmeier, Arnold

    2013-11-01

    We studied the interactions between the stellar wind plasma flow of a typical M star, such as GJ 436, and the hydrogen-rich upper atmosphere of an Earth-like planet and a "super-Earth" with a radius of 2 R(Earth) and a mass of 10 M(Earth), located within the habitable zone at ∼0.24 AU. We investigated the formation of extended atomic hydrogen coronae under the influences of the stellar XUV flux (soft X-rays and EUV), stellar wind density and velocity, shape of a planetary obstacle (e.g., magnetosphere, ionopause), and the loss of planetary pickup ions on the evolution of hydrogen-dominated upper atmospheres. Stellar XUV fluxes that are 1, 10, 50, and 100 times higher compared to that of the present-day Sun were considered, and the formation of high-energy neutral hydrogen clouds around the planets due to the charge-exchange reaction under various stellar conditions was modeled. Charge-exchange between stellar wind protons with planetary hydrogen atoms, and photoionization, lead to the production of initially cold ions of planetary origin. We found that the ion production rates for the studied planets can vary over a wide range, from ∼1.0×10²⁵ s⁻¹ to ∼5.3×10³⁰ s⁻¹, depending on the stellar wind conditions and the assumed XUV exposure of the upper atmosphere. Our findings indicate that most likely the majority of these planetary ions are picked up by the stellar wind and lost from the planet. Finally, we estimated the long-time nonthermal ion pickup escape for the studied planets and compared them with the thermal escape. According to our estimates, nonthermal escape of picked-up ionized hydrogen atoms over a planet's lifetime within the habitable zone of an M dwarf varies between ∼0.4 Earth ocean equivalent amounts of hydrogen (EO(H)) to <3 EO(H) and usually is several times smaller in comparison to the thermal atmospheric escape rates.

  11. Initial stages of oxidation of near-stoichiometric titanium carbide at low oxygen pressures

    International Nuclear Information System (INIS)

    Shabalin, I.L.; Vishnyakov, V.M.; Bull, D.J.; Keens, S.G.; Yamshchikov, L.F.; Shabalin, L.I.

    2009-01-01

    A novel approach to the oxidation mechanism of near-stoichiometric TiC is presented. It is confirmed by consideration of solid-state chemical kinetics model and electron microscopy observations in parallel. At low oxygen pressures and moderate temperatures the initial step of the process is connected with the dissolution of oxygen and subsequent decomposition of oxygen-oversaturated oxycarbide, which ultimately results in the nucleation of oxide phase, in particular anatase, belike stabilised by residual carbon. An anatase-rutile transformation is concurrent with deeper carbon burn-off in the oxide scale, which sinters at higher temperatures. This mechanism shifts the process to a gas diffusion regime, governed by the scale permeability, but determined by solid-state diffusion that is reflected in the kinetics, as further temperature increase is accompanied by a decrease of the oxidation rate, so in general the process is characterised by the negative value of apparent activation energy

  12. Methyl sterol and cyclopropane fatty acid composition of Methylococcus capsulatus grown at low oxygen tensions

    Science.gov (United States)

    Jahnke, L. L.; Nichols, P. D.

    1986-01-01

    The sterol and fatty acid concentrations for M. capsulatus grown in fed-batch cultures over a wide range of oxygen tensions (0.1-10.6 percent) and at a constant methane level are evaluated. The analyses reveal that the biomass decreases as oxygen levels are lowered; the sterol concentration increases when the oxygen range is between 0.5-1.1 percent and decreases when the oxygen range is below 0.5 percent; and the amount of monounsaturated C16 decreases and the concentration of cyclopropane fatty acids increases after oxygen is reduced. It is noted that growth and membrane synthesis occur at low oxygen concentrations and that the synthesis of membrane lipids responds to growth conditions.

  13. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  14. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  15. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  16. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  17. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  18. Low Oxygen Tension Enhances Expression of Myogenic Genes When Human Myoblasts Are Activated from G0 Arrest

    DEFF Research Database (Denmark)

    Sellathurai, Jeeva; Nielsen, Joachim; Hejbøl, Eva Kildall

    2016-01-01

    -PCR, immunocytochemistry and western blot. RESULTS AND CONCLUSIONS: We found an increase in proliferation rate of myoblasts when activated at a low oxygen tension (1% O2) compared to 21% O2. In addition, the gene expression studies showed up regulation of the myogenesis related genes PAX3, PAX7, MYOD, MYOG (myogenin), MET......, NCAM, DES (desmin), MEF2A, MEF2C and CDH15 (M-cadherin), however, the fraction of DES and MYOD positive cells was not increased by low oxygen tension, indicating that 1% O2 may not have a functional effect on the myogenic response. Furthermore, the expression of genes involved in the TGFβ, Notch...... and Wnt signaling pathways were also up regulated in low oxygen tension. The differences in gene expression were most pronounced at day one after activation from G0-arrest, thus the initial activation of myoblasts seemed most sensitive to changes in oxygen tension. Protein expression of HES1 and β...

  19. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  20. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

    Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

    1983-01-01

    Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

  1. Dependence of nitrite oxidation on nitrite and oxygen in low-oxygen seawater

    Science.gov (United States)

    Sun, Xin; Ji, Qixing; Jayakumar, Amal; Ward, Bess B.

    2017-08-01

    Nitrite oxidation is an essential step in transformations of fixed nitrogen. The physiology of nitrite oxidizing bacteria (NOB) implies that the rates of nitrite oxidation should be controlled by concentration of their substrate, nitrite, and the terminal electron acceptor, oxygen. The sensitivities of nitrite oxidation to oxygen and nitrite concentrations were investigated using 15N tracer incubations in the Eastern Tropical North Pacific. Nitrite stimulated nitrite oxidation under low in situ nitrite conditions, following Michaelis-Menten kinetics, indicating that nitrite was the limiting substrate. The nitrite half-saturation constant (Ks = 0.254 ± 0.161 μM) was 1-3 orders of magnitude lower than in cultivated NOB, indicating higher affinity of marine NOB for nitrite. The highest rates of nitrite oxidation were measured in the oxygen depleted zone (ODZ), and were partially inhibited by additions of oxygen. This oxygen sensitivity suggests that ODZ specialist NOB, adapted to low-oxygen conditions, are responsible for apparently anaerobic nitrite oxidation.

  2. Thermal stability of pulsed laser deposited iridium oxide thin films at low oxygen atmosphere

    Science.gov (United States)

    Gong, Yansheng; Wang, Chuanbin; Shen, Qiang; Zhang, Lianmeng

    2013-11-01

    Iridium oxide (IrO2) thin films have been regarded as a leading candidate for bottom electrode and diffusion barrier of ferroelectric capacitors, some process related issues need to be considered before integrating ferroelectric capacitors into memory cells. This paper presents the thermal stability of pulsed laser deposited IrO2 thin films at low oxygen atmosphere. Emphasis was given on the effect of post-deposition annealing temperature at different oxygen pressure (PO2) on the crystal structure, surface morphology, electrical resistivity, carrier concentration and mobility of IrO2 thin films. The results showed that the thermal stability of IrO2 thin films was strongly dependent on the oxygen pressure and annealing temperature. IrO2 thin films can stably exist below 923 K at PO2 = 1 Pa, which had a higher stability than the previous reported results. The surface morphology of IrO2 thin films depended on PO2 and annealing temperature, showing a flat and uniform surface for the annealed films. Electrical properties were found to be sensitive to both the annealing temperature and oxygen pressure. The room-temperature resistivity of IrO2 thin films with a value of 49-58 μΩ cm increased with annealing temperature at PO2 = 1 Pa. The thermal stability of IrO2 thin films as a function of oxygen pressure and annealing temperature was almost consistent with thermodynamic calculation.

  3. Dynamical Characterization of a Low Oxygen Submesoscale Coherent Vortex in the Eastern North Atlantic Ocean

    Science.gov (United States)

    Pietri, A.; Karstensen, J.

    2018-03-01

    A submesoscale coherent vortex (SCV) with a low oxygen core is characterized from underwater glider and mooring observations from the eastern tropical North Atlantic, north of the Cape Verde Islands. The eddy crossed the mooring with its center and a 1 month time series of the SCV's hydrographic and upper 100 m currents structure was obtained. About 45 days after, and ˜100 km west, the SCV frontal zone was surveyed in high temporal and spatial resolution using an underwater glider. Satellite altimetry showed the SCV was formed about 7 months before at the Mauritanian coast. The SCV was located at 80-100 m depth, its diameter was ˜100 km and its maximum swirl velocity ˜0.4 m s-1. A Burger number of 0.2 and a vortex Rossby number 0.15 indicate a flat lens in geostrophic balance. Mooring and glider data show in general comparable dynamical and thermohaline structures, the glider in high spatial resolution, the mooring in high temporal resolution. Surface maps of chlorophyll concentration suggest high productivity inside and around the SCV. The low potential vorticity (PV) core of the SCV is surrounded by filamentary structures, sloping down at different angles from the mixed layer base and with typical width of 10-20 km and a vertical extent of 50-100 m.

  4. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  5. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  6. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

    Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

  7. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  8. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

    Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

  9. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  10. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  11. Marine species in ambient low-oxygen regions subject to double jeopardy impacts of climate change.

    Science.gov (United States)

    Stortini, Christine H; Chabot, Denis; Shackell, Nancy L

    2017-06-01

    We have learned much about the impacts of warming on the productivity and distribution of marine organisms, but less about the impact of warming combined with other environmental stressors, including oxygen depletion. Also, the combined impact of multiple environmental stressors requires evaluation at the scales most relevant to resource managers. We use the Gulf of St. Lawrence, Canada, characterized by a large permanently hypoxic zone, as a case study. Species distribution models were used to predict the impact of multiple scenarios of warming and oxygen depletion on the local density of three commercially and ecologically important species. Substantial changes are projected within 20-40 years. A eurythermal depleted species already limited to shallow, oxygen-rich refuge habitat (Atlantic cod) may be relatively uninfluenced by oxygen depletion but increase in density within refuge areas with warming. A more stenothermal, deep-dwelling species (Greenland halibut) is projected to lose ~55% of its high-density areas under the combined impacts of warming and oxygen depletion. Another deep-dwelling, more eurythermal species (Northern shrimp) would lose ~4% of its high-density areas due to oxygen depletion alone, but these impacts may be buffered by warming, which may increase density by 8% in less hypoxic areas, but decrease density by ~20% in the warmest parts of the region. Due to local climate variability and extreme events, and that our models cannot project changes in species sensitivity to hypoxia with warming, our results should be considered conservative. We present an approach to effectively evaluate the individual and cumulative impacts of multiple environmental stressors on a species-by-species basis at the scales most relevant to managers. Our study may provide a basis for work in other low-oxygen regions and should contribute to a growing literature base in climate science, which will continue to be of support for resource managers as climate change

  12. Endogenous adaptation to low oxygen modulates T-cell regulatory pathways in EAE.

    Science.gov (United States)

    Esen, Nilufer; Katyshev, Vladimir; Serkin, Zakhar; Katysheva, Svetlana; Dore-Duffy, Paula

    2016-01-19

    In the brain, chronic inflammatory activity may lead to compromised delivery of oxygen and glucose suggesting that therapeutic approaches aimed at restoring metabolic balance may be useful. In vivo exposure to chronic mild normobaric hypoxia (10 % oxygen) leads to a number of endogenous adaptations that includes vascular remodeling (angioplasticity). Angioplasticity promotes tissue survival. We have previously shown that induction of adaptive angioplasticity modulates the disease pattern in myelin oligodendrocyte glycoprotein (MOG)-induced experimental autoimmune encephalomyelitis (EAE). In the present study, we define mechanisms by which adaptation to low oxygen functionally ameliorates the signs and symptoms of EAE and for the first time show that tissue hypoxia may fundamentally alter neurodegenerative disease. C57BL/6 mice were immunized with MOG, and some of them were kept in the hypoxia chambers (day 0) and exposed to 10 % oxygen for 3 weeks, while the others were kept at normoxic environment. Sham-immunized controls were included in both hypoxic and normoxic groups. Animals were sacrificed at pre-clinical and peak disease periods for tissue collection and analysis. Exposure to mild hypoxia decreased histological evidence of inflammation. Decreased numbers of cluster of differentiation (CD)4+ T cells were found in the hypoxic spinal cords associated with a delayed Th17-specific cytokine response. Hypoxia-induced changes did not alter the sensitization of peripheral T cells to the MOG peptide. Exposure to mild hypoxia induced significant increases in anti-inflammatory IL-10 levels and an increase in the number of spinal cord CD25+FoxP3+ T-regulatory cells. Acclimatization to mild hypoxia incites a number of endogenous adaptations that induces an anti-inflammatory milieu. Further understanding of these mechanisms system may pinpoint possible new therapeutic targets to treat neurodegenerative disease.

  13. Diesel autothermal reforming with hydrogen peroxide for low-oxygen environments

    International Nuclear Information System (INIS)

    Han, Gwangwoo; Lee, Sangho; Bae, Joongmyeon

    2015-01-01

    Highlights: • The concept of diesel reforming using hydrogen peroxide was newly proposed. • Characteristics of hydrogen peroxide was experimentally investigated. • Thermodynamically possible operating conditions were analyzed. • Catalytic performance of Ni–Ru/CGO for various diesel compounds was evaluated. • Long-term testing was successfully conducted using Korean commercial diesel. - Abstract: To operate fuel cells effectively in low-oxygen environments, such as in submarines and unmanned underwater vehicles, a hydrogen source with high hydrogen storage density is required. In this paper, diesel autothermal reforming (ATR) with hydrogen peroxide as an alternative oxidant is proposed as a hydrogen production method. Diesel fuel has higher hydrogen density than metal hydrides or other hydrocarbons. In addition, hydrogen peroxide can decompose into steam and oxygen, which are required for diesel ATR. Moreover, both diesel fuel and hydrogen peroxide are liquid states, enabling easy storage for submarine applications. Hydrogen peroxide exhibited the same characteristics as steam and oxygen when used as an oxidant in diesel reforming when pre-decomposition method was used. The thermodynamically calculated operating conditions were a steam-to-carbon ratio (SCR) of 3.0, an oxygen-to-carbon ratio (OCR) of 0.5, and temperatures below 700 °C to account for safety issues associated with hydrogen peroxide use and exothermic reactions. Catalytic activity and stability tests over Ni–Ru (19.5–0.5 wt.%)/Ce 0.9 Gd 0.1 O 2−x were conducted using various diesel compounds. Furthermore, long-term diesel ATR tests were conducted for 200 h using Korean commercial diesel. The degradation rate was 3.67%/100 h without the production of ethylene

  14. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  15. Study on combustion characteristics of dimethyl ether under the moderate or intense low-oxygen dilution condition

    International Nuclear Information System (INIS)

    Kang, Yinhu; Lu, Tianfeng; Lu, Xiaofeng; Wang, Quanhai; Huang, Xiaomei; Peng, Shini; Yang, Dong; Ji, Xuanyu; Song, Yangfan

    2016-01-01

    Highlights: • Oxygen content in the flame base increased due to the prolonged ignition delay time. • Flow field in the furnace affected thermal/chemical structure of the flame partially. • Preheating and dilution facilitated moderate or intense low-oxygen dilution regime. • Dominant pollutant formation ways of dimethyl ether in hot dilution were clarified. • Preheating and dilution reduced nitrogen oxide emission of dimethyl ether. - Abstract: Experiments and numerical simulations were conducted in this paper to study the combustion behavior of dimethyl ether in the moderate or intense low-oxygen dilution regime, in terms of thermal/chemical structure and chemical kinetics associated with nitrogen oxide and carbon monoxide emissions. Several co-flow temperatures and oxygen concentrations were involved in the experiments to investigate their impacts on the flame behavior systematically. The results show that in the moderate or intense low-oxygen dilution regime, oxygen concentrations in the flame base slightly increased because of the prolonged ignition delay time of the reactant mixture due to oxidizer dilution, which changed the local combustion process and composition considerably. The oxidation rates of hydrocarbons were significantly depressed in the moderate or intense low-oxygen dilution regime, such that a fraction of unburned hydrocarbons at the furnace outlet were recirculated into the outer annulus of the furnace, which changed the local radial profiles of carbon monoxide, methane, and hydrogen partially. Moreover, with the increment in co-flow temperature or oxygen mole fraction, flame temperature, and hydroxyl radical, carbon monoxide, and hydrogen mole fractions across the reaction zone increased gradually. For the dimethyl ether-moderate or intense low-oxygen dilution flame, temperature homogeneity was improved at higher co-flow temperature or lower oxygen mole fraction. The carbon monoxide emission depended on the levels of temperature and

  16. The role of biology in planetary evolution: cyanobacterial primary production in low-oxygen Proterozoic oceans.

    Science.gov (United States)

    Hamilton, Trinity L; Bryant, Donald A; Macalady, Jennifer L

    2016-02-01

    Understanding the role of biology in planetary evolution remains an outstanding challenge to geobiologists. Progress towards unravelling this puzzle for Earth is hindered by the scarcity of well-preserved rocks from the Archean (4.0 to 2.5 Gyr ago) and Proterozoic (2.5 to 0.5 Gyr ago) Eons. In addition, the microscopic life that dominated Earth's biota for most of its history left a poor fossil record, consisting primarily of lithified microbial mats, rare microbial body fossils and membrane-derived hydrocarbon molecules that are still challenging to interpret. However, it is clear from the sulfur isotope record and other geochemical proxies that the production of oxygen or oxidizing power radically changed Earth's surface and atmosphere during the Proterozoic Eon, pushing it away from the more reducing conditions prevalent during the Archean. In addition to ancient rocks, our reconstruction of Earth's redox evolution is informed by our knowledge of biogeochemical cycles catalysed by extant biota. The emergence of oxygenic photosynthesis in ancient cyanobacteria represents one of the most impressive microbial innovations in Earth's history, and oxygenic photosynthesis is the largest source of O2 in the atmosphere today. Thus the study of microbial metabolisms and evolution provides an important link between extant biota and the clues from the geologic record. Here, we consider the physiology of cyanobacteria (the only microorganisms capable of oxygenic photosynthesis), their co-occurrence with anoxygenic phototrophs in a variety of environments and their persistence in low-oxygen environments, including in water columns as well as mats, throughout much of Earth's history. We examine insights gained from both the rock record and cyanobacteria presently living in early Earth analogue ecosystems and synthesize current knowledge of these ancient microbial mediators in planetary redox evolution. Our analysis supports the hypothesis that anoxygenic photosynthesis

  17. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  18. The influence of low oxygen and contaminated sodium environments on the fatigue behavior of solution treated AISI 316 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, P [CEGB, BNL, Berkeley (United Kingdom)

    1977-07-01

    The influence of air and sodium environments on the fatigue properties of solution treated AISI 316 steel was studied by predictive methods and by conducting tests in air, in high temperature sodium, or following pre-exposure to sodium. The sodium environments studied included contaminated sodium or the products of sodium/water flames possibly typical of fast reactor fault conditions, and low oxygen sodium more appropriate to normal plant operation. Generally, fatigue properties were reduced by contaminated sodium or the products of sodium/water flames and improved by low oxygen sodium when compared with similar tests conducted in air. However, complex effects were observed with respect to crack initiation. The experimental results are discussed and generally follow trends predicted by physically based fatigue models. (author)

  19. Impact of low oxygen tension on stemness, proliferation and differentiation potential of human adipose-derived stem cells

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jane Ru; Pingguan-Murphy, Belinda; Wan Abas, Wan Abu Bakar [Department of Biomedical Engineering, Faculty of Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Noor Azmi, Mat Adenan; Omar, Siti Zawiah [Department of Obstetrics and Gynaecology, Faculty of Medicine, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Chua, Kien Hui [Department of Physiology, Faculty of Medicine, Universiti Kebangsaan Malaysia, Jalan Raja Muda Abdul Aziz, 50300 Kuala Lumpur (Malaysia); Wan Safwani, Wan Kamarul Zaman, E-mail: wansafwani@um.edu.my [Department of Biomedical Engineering, Faculty of Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia)

    2014-05-30

    Highlights: • Hypoxia maintains the stemness of adipose-derived stem cells (ASCs). • ASCs show an increased proliferation rate under low oxygen tension. • Oxygen level as low as 2% enhances the chondrogenic differentiation potential of ASCs. • HIF-1α may regulate the proliferation and differentiation activities of ASCs under hypoxia. - Abstract: Adipose-derived stem cells (ASCs) have been found adapted to a specific niche with low oxygen tension (hypoxia) in the body. As an important component of this niche, oxygen tension has been known to play a critical role in the maintenance of stem cell characteristics. However, the effect of O{sub 2} tension on their functional properties has not been well determined. In this study, we investigated the effects of O{sub 2} tension on ASCs stemness, differentiation and proliferation ability. Human ASCs were cultured under normoxia (21% O{sub 2}) and hypoxia (2% O{sub 2}). We found that hypoxia increased ASC stemness marker expression and proliferation rate without altering their morphology and surface markers. Low oxygen tension further enhances the chondrogenic differentiation ability, but reduces both adipogenic and osteogenic differentiation potential. These results might be correlated with the increased expression of HIF-1α under hypoxia. Taken together, we suggest that growing ASCs under 2% O{sub 2} tension may be important in expanding ASCs effectively while maintaining their functional properties for clinical therapy, particularly for the treatment of cartilage defects.

  20. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  1. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  2. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  3. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.

    1985-01-01

    The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

  4. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  5. Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

    2001-01-01

    In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

  6. On-Line Coolant Chemistry Analysis

    International Nuclear Information System (INIS)

    LM Bachman

    2006-01-01

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

  7. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  8. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  9. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  10. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  11. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  12. Low Oxygen Modulates Multiple Signaling Pathways, Increasing Self-Renewal, While Decreasing Differentiation, Senescence, and Apoptosis in Stromal MIAMI Cells

    Science.gov (United States)

    Rios, Carmen; D'Ippolito, Gianluca; Curtis, Kevin M.; Delcroix, Gaëtan J.-R.; Gomez, Lourdes A.; El Hokayem, Jimmy; Rieger, Megan; Parrondo, Ricardo; de las Pozas, Alicia; Perez-Stable, Carlos; Howard, Guy A.

    2016-01-01

    Human bone marrow multipotent mesenchymal stromal cell (hMSC) number decreases with aging. Subpopulations of hMSCs can differentiate into cells found in bone, vasculature, cartilage, gut, and other tissues and participate in their repair. Maintaining throughout adult life such cell subpopulations should help prevent or delay the onset of age-related degenerative conditions. Low oxygen tension, the physiological environment in progenitor cell-rich regions of the bone marrow microarchitecture, stimulates the self-renewal of marrow-isolated adult multilineage inducible (MIAMI) cells and expression of Sox2, Nanog, Oct4a nuclear accumulation, Notch intracellular domain, notch target genes, neuronal transcriptional repressor element 1 (RE1)-silencing transcription factor (REST), and hypoxia-inducible factor-1 alpha (HIF-1α), and additionally, by decreasing the expression of (i) the proapoptotic proteins, apoptosis-inducing factor (AIF) and Bak, and (ii) senescence-associated p53 expression and β-galactosidase activity. Furthermore, low oxygen increases canonical Wnt pathway signaling coreceptor Lrp5 expression, and PI3K/Akt pathway activation. Lrp5 inhibition decreases self-renewal marker Sox2 mRNA, Oct4a nuclear accumulation, and cell numbers. Wortmannin-mediated PI3K/Akt pathway inhibition leads to increased osteoblastic differentiation at both low and high oxygen tension. We demonstrate that low oxygen stimulates a complex signaling network involving PI3K/Akt, Notch, and canonical Wnt pathways, which mediate the observed increase in nuclear Oct4a and REST, with simultaneous decrease in p53, AIF, and Bak. Collectively, these pathway activations contribute to increased self-renewal with concomitant decreased differentiation, cell cycle arrest, apoptosis, and/or senescence in MIAMI cells. Importantly, the PI3K/Akt pathway plays a central mechanistic role in the oxygen tension-regulated self-renewal versus osteoblastic differentiation of progenitor cells. PMID:27059084

  13. Revised Mark 22 coolant temperature coefficients

    International Nuclear Information System (INIS)

    Graves, W.E.

    1987-01-01

    Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

  14. Freeform Deposition Method for Coolant Channel Closeout

    Science.gov (United States)

    Gradl, Paul R. (Inventor); Reynolds, David Christopher (Inventor); Walker, Bryant H. (Inventor)

    2017-01-01

    A method is provided for fabricating a coolant channel closeout jacket on a structure having coolant channels formed in an outer surface thereof. A line of tangency relative to the outer surface is defined for each point on the outer surface. Linear rows of a metal feedstock are directed towards and deposited on the outer surface of the structure as a beam of weld energy is directed to the metal feedstock so-deposited. A first angle between the metal feedstock so-directed and the line of tangency is maintained in a range of 20-90.degree.. The beam is directed towards a portion of the linear rows such that less than 30% of the cross-sectional area of the beam impinges on a currently-deposited one of the linear rows. A second angle between the beam and the line of tangency is maintained in a range of 5-65 degrees.

  15. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  16. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  17. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  18. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  19. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  20. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  1. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  2. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  3. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  4. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  5. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  6. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  7. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  8. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  9. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  10. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  11. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

    Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

  12. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  13. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  14. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  15. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  16. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  17. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  18. CFD analyses of coolant channel flowfields

    Science.gov (United States)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  19. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  20. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

  1. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  2. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  3. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  4. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  5. Antioxidant effect of aromatic volatiles emitted by Lavandula dentata, Mentha spicata, and M. piperita on mouse subjected to low oxygen condition.

    Science.gov (United States)

    Hu, Zenghui; Wang, Chunling; Shen, Hong; Zhang, Kezhong; Leng, Pingsheng

    2017-12-01

    This study aims to investigate the antioxidant effect of aromatic volatiles of three common aromatic plants, Lavandula dentata, Mentha spicata, and M. piperita. In this study, kunming mice subjected to low oxygen condition were treated with the volatiles emitted from these aromatic plants through inhalation administration. Then the blood cell counts, and the activities and gene expressions of antioxidant enzymes in different tissues were tested. The results showed that low oxygen increased the counts of red blood cells, white blood cells, and blood platelets of mice, and aromatic volatiles decreased their counts. Exposure to aromatic volatiles resulted in decreases in the malonaldehyde contents, and increases in the activities and gene expressions of superoxide dismutase, glutathione peroxidase, and catalase in different tissues under low oxygen. In addition, as the main component of aromatic volatiles, eucalyptol was the potential source that imparted positive antioxidant effect.

  6. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  7. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  8. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  9. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  10. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  11. Reactor coolant flow measurements at Point Lepreau

    International Nuclear Information System (INIS)

    Brenciaglia, G.; Gurevich, Y.; Liu, G.

    1996-01-01

    The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

  12. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  13. Reactive oxygen species-driven HIF1α triggers accelerated glycolysis in endothelial cells exposed to low oxygen tension

    International Nuclear Information System (INIS)

    Paik, Jin-Young; Jung, Kyung-Ho; Lee, Jin-Hee; Park, Jin-Won; Lee, Kyung-Han

    2017-01-01

    Endothelial cells and their metabolic state regulate glucose transport into underlying tissues. Here, we show that low oxygen tension stimulates human umbilical vein endothelial cell 18 F–fluorodeoxyglucose ( 18 F–FDG) uptake and lactate production. This was accompanied by augmented hexokinase activity and membrane Glut-1, and increased accumulation of hypoxia-inducible factor-1α (HIF1α). Restoration of oxygen reversed the metabolic effect, but this was blocked by HIF1α stabilization. Hypoxia-stimulated 18 F–FDG uptake was completely abrogated by silencing of HIF1α expression or by a specific inhibitor. There was a rapid and marked increase of reactive oxygen species (ROS) by hypoxia, and ROS scavenging or NADPH oxidase inhibition completely abolished hypoxia-stimulated HIF1α and 18 F–FDG accumulation, placing ROS production upstream of HIF1α signaling. Hypoxia-stimulated HIF1α and 18 F–FDG accumulation was blocked by the protein kinase C (PKC) inhibitor, staurosporine. The phosphatidylinositol 3-kinase (PI3K) inhibitor, wortmannin, blocked hypoxia-stimulated 18 F–FDG uptake and attenuated hypoxia-responsive element binding of HIF1α without influencing its accumulation. Thus, ROS-driven HIF1α accumulation, along with PKC and PI3K signaling, play a key role in triggering accelerated glycolysis in endothelial cells under hypoxia, thereby contributing to 18 F–FDG transport.

  14. Zooplankton Responses to Low-Oxygen Condition upon a Shallow Oxygen Minimum Zone in the Upwelling Region off Chile

    Science.gov (United States)

    Hidalgo, P.; Escribano, R.

    2015-12-01

    A shallow oxygen minimum zone (OMZ) is a critical component in the coastal upwelling ecosystem off Chile. This OMZ causes oxygen-deficient water entering the photic layer and affecting plankton communities having low tolerance to hypoxia. Variable, and usually species-dependent, responses of zooplankton to hypoxia condition can be found. Most dominant species avoid hypoxia by restricting their vertical distribution, while others can temporarily enter and even spent part of their life cycle within the OMZ. Whatever the case, low-oxygen conditions appear to affect virtually all vital rates of zooplankton, such as mortality, fecundity, development and growth and metabolism, and early developmental stages seem more sensitive, with significant consequences for population and community dynamics. For most study cases, these effects are negative at individual and population levels. Observations and predictions upon increasing upwelling intensity over the last 20-30 years indicate a gradual shoaling of the OMZ, and so that an expected enhancement of these negative effects of hypoxia on the zooplankton community. Unknown processes of adaptation and community-structure adjustments are expected to take place with uncertain consequences for the food web of this highly productive eastern boundary current ecosystem.

  15. Regeneration of volatile compounds in Fuji apples following ultra low oxygen atmosphere storage and its effect on sensory acceptability.

    Science.gov (United States)

    Altisent, Rosa; Graell, Jordi; Lara, Isabel; López, Luisa; Echeverría, Gemma

    2008-09-24

    The aim of this work was to assess whether extra time spent under AIR conditions after storage in an ultra low oxygen (ULO) atmosphere could allow the regeneration of volatile compound emission without negatively affecting quality parameters and the consumer acceptability of Fuji apples. Fruits were stored for 19 and 30 weeks at 1 degrees C and 92% RH under ULO atmosphere conditions (1 kPa O 2:1 kPa CO 2) or under ULO conditions followed by different periods (2 and 4 weeks) in cold AIR atmosphere (ULO + 2w or ULO + 4w, respectively). Standard quality and emission of volatile compounds were analyzed after storage plus 1 and 7 days at 20 degrees C. Sensory attributes and acceptability were also determined after 7 days at 20 degrees C. The extra period of 30 weeks in an AIR atmosphere after ULO storage resulted in an increase in the concentration of the compounds that most contribute to the flavor of Fuji apples. These fruits were relatively well accepted by consumers despite a slight decline in firmness and acidity.

  16. Low oxygen affects photophysiology and the level of expression of two-carbon metabolism genes in the seagrass Zostera muelleri.

    Science.gov (United States)

    Kim, Mikael; Brodersen, Kasper Elgetti; Szabó, Milán; Larkum, Anthony W D; Raven, John A; Ralph, Peter J; Pernice, Mathieu

    2018-05-01

    Seagrasses are a diverse group of angiosperms that evolved to live in shallow coastal waters, an environment regularly subjected to changes in oxygen, carbon dioxide and irradiance. Zostera muelleri is the dominant species in south-eastern Australia, and is critical for healthy coastal ecosystems. Despite its ecological importance, little is known about the pathways of carbon fixation in Z. muelleri and their regulation in response to environmental changes. In this study, the response of Z. muelleri exposed to control and very low oxygen conditions was investigated by using (i) oxygen microsensors combined with a custom-made flow chamber to measure changes in photosynthesis and respiration, and (ii) reverse transcription quantitative real-time PCR to measure changes in expression levels of key genes involved in C 4 metabolism. We found that very low levels of oxygen (i) altered the photophysiology of Z. muelleri, a characteristic of C 3 mechanism of carbon assimilation, and (ii) decreased the expression levels of phosphoenolpyruvate carboxylase and carbonic anhydrase. These molecular-physiological results suggest that regulation of the photophysiology of Z. muelleri might involve a close integration between the C 3 and C 4 , or other CO 2 concentrating mechanisms metabolic pathways. Overall, this study highlights that the photophysiological response of Z. muelleri to changing oxygen in water is capable of rapid acclimation and the dynamic modulation of pathways should be considered when assessing seagrass primary production.

  17. Regulation of H+ Extrusion and Cytoplasmic pH in Maize Root Tips Acclimated to a Low-Oxygen Environment.

    Science.gov (United States)

    Xia, J. H.; Roberts, JKM.

    1996-05-01

    We tested the hypothesis that H+ extrusion contributes to cytoplasmic pH regulation and tolerance of anoxia in maize (Zea mays) root tips. We studied root tips of whole seedlings that were acclimated to a low-oxygen environment by pretreatment in 3% (v/v) O2. Acclimated root tips characteristically regulate cytoplasmic pH near neutrality and survive prolonged anoxia, whereas nonacclimated tips undergo severe cytoplasmic acidosis and die much more quickly. We show that the plasma membrane H+-ATPase can operate under anoxia and that net H+ extrusion increases when cytoplasmic pH falls. However, at an external pH near 6.0, H+ extrusion contributes little to cytoplasmic pH regulation. At more acidic external pH values, net H+ flux into root tips increases dramatically, leading to a decrease in cytoplasmic pH and reduced tolerance of anoxia. We present evidence that, under these conditions, H+ pumps are activated to partly offset acidosis due to H+ influx and, thereby, contribute to cytoplasmic pH regulation and tolerance of anoxia. The regulation of H+ extrusion under anoxia is discussed with respect to the acclimation response and mechanisms of intracellular pH regulation in aerobic plant cells.

  18. Coolant rate distribution in horizontal steam generator under natural circulation

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered

  19. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  20. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  1. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  2. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

    In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

  3. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A; Leontieva, V; Mitrioukhin, A [St. Petersburg State Technical Univ. (Russian Federation)

    1998-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  4. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  5. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  6. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  7. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  8. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  9. Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

    1976-04-01

    The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

  10. Hydrogen-Rich Saline Attenuates Brain Injury Induced by Cardiopulmonary Bypass and Inhibits Microvascular Endothelial Cell Apoptosis Via the PI3K/Akt/GSK3β Signaling Pathway in Rats

    Directory of Open Access Journals (Sweden)

    Keyan Chen

    2017-10-01

    Full Text Available Background/Aims: Cardiopulmonary bypass (CPB is prone to inducing brain injury during open heart surgery. A hydrogen-rich solution (HRS can prevent oxidation and apoptosis, and inhibit inflammation. This study investigated effects of HRS on brain injury induced by CPB and regulatory mechanisms of the PI3K/Akt/GSK3β signaling pathway. Methods: A rat CPB model and an in vitro cell hypoxia model were established. After HRS treatment, Rat behavior was measured using neurological deficit score; Evans blue (EB was used to assess permeability of the blood-brain barrier (BBB; HE staining was used to observe pathological changes; Inflammatory factors and brain injury markers were detected by ELISA; the PI3K/Akt/GSK3β pathway-related proteins and apoptosis were assessed by western blot, immunohistochemistry and qRT –PCR analyses of brain tissue and neurons. Results: After CPB, brain tissue anatomy was disordered, and cell structure was abnormal. Brain tissue EB content increased. There was an increase in the number of apoptotic cells, an increase in expression of Bax and caspase-3, a decrease in expression of Bcl2, and increases in levels of Akt, GSK3β, P-Akt, and P-GSK3β in brain tissue. HRS treatment attenuated the inflammatory reaction ,brain tissue EB content was significantly reduced and significantly decreased expression levels of Bax, caspase-3, Akt, GSK3β, P-Akt, and P-GSK3β in the brain. After adding the PI3K signaling pathway inhibitor, LY294002, to rat cerebral microvascular endothelial cells (CMECs, HRS could reduce activated Akt expression and downstream regulatory gene phosphorylation of GSK3β expression, and inhibit CMEC apoptosis. Conclusion: The PI3K/Akt/GSK3β signaling pathway plays an important role in the mechanism of CPB-induced brain injury. HRS can reduce CPB-induced brain injury and inhibit CMEC apoptosis through the PI3K/Akt/GSK3β signaling pathway.

  11. Gene expression and physiological changes of different populations of the long-lived bivalve Arctica islandica under low oxygen conditions.

    Directory of Open Access Journals (Sweden)

    Eva E R Philipp

    Full Text Available The bivalve Arctica islandica is extremely long lived (>400 years and can tolerate long periods of hypoxia and anoxia. European populations differ in maximum life spans (MLSP from 40 years in the Baltic to >400 years around Iceland. Characteristic behavior of A. islandica involves phases of metabolic rate depression (MRD during which the animals burry into the sediment for several days. During these phases the shell water oxygen concentrations reaches hypoxic to anoxic levels, which possibly support the long life span of some populations. We investigated gene regulation in A. islandica from a long-lived (MLSP 150 years German Bight population and the short-lived Baltic Sea population, experimentally exposed to different oxygen levels. A new A. islandica transcriptome enabled the identification of genes important during hypoxia/anoxia events and, more generally, gene mining for putative stress response and (anti- aging genes. Expression changes of a antioxidant defense: Catalase, Glutathione peroxidase, manganese and copper-zinc Superoxide dismutase; b oxygen sensing and general stress response: Hypoxia inducible factor alpha, Prolyl hydroxylase and Heat-shock protein 70; and c anaerobic capacity: Malate dehydrogenase and Octopine dehydrogenase, related transcripts were investigated. Exposed to low oxygen, German Bight individuals suppressed transcription of all investigated genes, whereas Baltic Sea bivalves enhanced gene transcription under anoxic incubation (0 kPa and, further, decreased these transcription levels again during 6 h of re-oxygenation. Hypoxic and anoxic exposure and subsequent re-oxygenation in Baltic Sea animals did not lead to increased protein oxidation or induction of apoptosis, emphasizing considerable hypoxia/re-oxygenation tolerance in this species. The data suggest that the energy saving effect of MRD may not be an attribute of Baltic Sea A. islandica chronically exposed to high environmental variability (oxygenation

  12. Electrochemical Exfoliation of Graphite in Aqueous Sodium Halide Electrolytes toward Low Oxygen Content Graphene for Energy and Environmental Applications.

    Science.gov (United States)

    Munuera, J M; Paredes, J I; Enterría, M; Pagán, A; Villar-Rodil, S; Pereira, M F R; Martins, J I; Figueiredo, J L; Cenis, J L; Martínez-Alonso, A; Tascón, J M D

    2017-07-19

    Graphene and graphene-based materials have shown great promise in many technological applications, but their large-scale production and processing by simple and cost-effective means still constitute significant issues in the path of their widespread implementation. Here, we investigate a straightforward method for the preparation of a ready-to-use and low oxygen content graphene material that is based on electrochemical (anodic) delamination of graphite in aqueous medium with sodium halides as the electrolyte. Contrary to previous conflicting reports on the ability of halide anions to act as efficient exfoliating electrolytes in electrochemical graphene exfoliation, we show that proper choice of both graphite electrode (e.g., graphite foil) and sodium halide concentration readily leads to the generation of large quantities of single-/few-layer graphene nanosheets possessing a degree of oxidation (O/C ratio down to ∼0.06) lower than that typical of anodically exfoliated graphenes obtained with commonly used electrolytes. The halide anions are thought to play a role in mitigating the oxidation of the graphene lattice during exfoliation, which is also discussed and rationalized. The as-exfoliated graphene materials exhibited a three-dimensional morphology that was suitable for their practical use without the need to resort to any kind of postproduction processing. When tested as dye adsorbents, they outperformed many previously reported graphene-based materials (e.g., they adsorbed ∼920 mg g -1 for methyl orange) and were useful sorbents for oils and nonpolar organic solvents. Supercapacitor cells assembled directly from the as-exfoliated products delivered energy and power density values (up to 15.3 Wh kg -1 and 3220 W kg -1 , respectively) competitive with those of many other graphene-based devices but with the additional advantage of extreme simplicity of preparation.

  13. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  14. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  15. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  16. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

    Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

  17. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

    Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

    2004-01-01

    Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

  18. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  19. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  20. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  1. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  2. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  3. Speed control device for coolant recycling pump

    International Nuclear Information System (INIS)

    Kageyama, Takao.

    1992-01-01

    The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

  4. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  5. Reactor Coolant Temperature Measurement using Ultrasonic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, JaeCheon [KEPCO International Nuclear graduate School, Ulsan (Korea, Republic of); Seo, YongSun; Bechue, Nicholas [Krohne Messtechnik GmbH, Duisburg (Germany)

    2016-10-15

    In NPP, the primary piping temperature is detected by four redundant RTDs (Resistance Temperature Detectors) installed 90 degrees apart on the RCS (Reactor Coolant System) piping circumferentially. Such outputs however, if applied to I and C systems would not give balanced results. The discrepancy can be explained by either thermal stratification or improper arrangement of thermo-wells and RTDs. This phenomenon has become more pronounced in the hot-leg piping than in the cold-leg. Normally, the temperature difference among channels is in the range of 1°F in Korean nuclear power Plants. Consequently, a more accurate pipe average temperate measurement technique is required. Ultrasonic methods can be used to measure average temperatures with relatively higher accuracy than RTDs because the sound wave propagation in the RCS pipe is proportional to the average temperature around pipe area. The inaccuracy of RCS temperature measurement worsens the safety margin for both DNBR and LPD. The possibility of this discrepancy has been reported with thermal stratification effect. Proposed RCS temperature measurement system based on ultrasonic technology offers a countermeasure to cope with thermal stratification effect on hot-leg piping that has been an unresolved issue in NPPs. By introducing ultrasonic technology, the average internal piping temperature can be measured with high accuracy. The inaccuracy can be decreased less than ±1℉ by this method.

  6. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  7. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  8. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  9. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-11-01

    A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

  10. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  11. Evaluation of filtration and distillation methods for recycling automotive coolant

    International Nuclear Information System (INIS)

    Randall, P.M.; Gavaskar, A.R.

    1992-01-01

    Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

  12. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  13. IVF/ICSI outcomes after culture of human embryos at low oxygen tension: a meta-analysis

    Directory of Open Access Journals (Sweden)

    Gomes Sobrinho David B

    2011-11-01

    different between the groups. Conclusions Despite some promising results, it seems too early to conclude that low O2 culture has an effect on IVF outcome. Additional randomised controlled trials are necessary before evidence-based recommendations can be provided. It should be emphasised that the present meta-analysis does not provide any evidence that low oxygen concentration is unnecessary.

  14. Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

    Science.gov (United States)

    Woon, K. S.; Tnay, G. L.; Rahman, M.

    2018-05-01

    Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

  15. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  16. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  17. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  18. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  19. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  20. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  1. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  2. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  3. The effects of intermittent exposure to low-pH and low-oxygen conditions on survival and growth of juvenile red abalone

    Science.gov (United States)

    Kim, T. W.; Barry, J. P.; Micheli, F.

    2013-11-01

    Exposure of nearshore animals to hypoxic, low-pH waters upwelled from below the continental shelf and advected near the coast may be stressful to marine organisms and lead to impaired physiological performance. We mimicked upwelling conditions in the laboratory and tested the effect of fluctuating exposure to water with low-pH and/or low-oxygen levels on the mortality and growth of juvenile red abalone (Haliotis rufescens, shell length 5-10 mm). Mortality rates of juvenile abalone exposed to low-pH (7.5, total scale) and low-O2 (40% saturation, mg L-1) conditions for periods of 3 to 6 h every 3-5 days over 2 weeks did not differ from those exposed to control conditions (O2: 100% saturation, 12 mg L-1; pH 8.0). However, when exposure was extended to 24 h, twice over a 15-day period, juveniles experienced 5-20% higher mortality in the low-oxygen treatments compared to control conditions. Growth rates were reduced significantly when juveniles were exposed to low-oxygen and low-pH treatments. Furthermore, individual variation of growth rate increased when juveniles were exposed simultaneously to low-pH and low-O2 conditions. These results indicate that prolonged exposure to low-oxygen levels is detrimental for the survival of red abalone, whereas pH is a crucial factor for their growth. However, the high individual variation in growth rate under low levels of both pH and oxygen suggests that cryptic phenotypic plasticity may promote resistance to prolonged upwelling conditions by a portion of the population.

  4. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  5. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  6. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  7. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  8. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  9. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  10. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  11. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  12. Actively controlling coolant-cooled cold plate configuration

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  13. Sloshing of coolant in a seismically isolated reactor

    International Nuclear Information System (INIS)

    Wu, T.S.; Guildys, J.; Seidensticker, R.W.

    1988-01-01

    During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

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

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    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

  15. Design and development of remotely operated coolant channel cutting machine

    International Nuclear Information System (INIS)

    Suthar, R.L.; Sinha, A.K.; Srikrishnamurty, G.

    1994-01-01

    One of the coolant tubes of Narora Atomic Power Station (NAPS) reactor needs to be removed. To remove a coolant tube, four cutting operations, (liner tube cutting, end-fitting cutting, machining of seal weld of bellow ring and finally coolant tube cutting) are required to be carried out. A remotely operated cutting machine to carry out all these operations has been designed and developed by Central Workshops. This machine is able to cut at the exact location because of numerically controlled axial and radial travel of tool. Only by changing the tool head and tool holder, same machine can be used for various types of cutting/machining operations. This report details the design, manufacture, assembly and testing work done on the machine. (author). 4 figs

  16. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  17. Method of eliminating cruds in the primary coolants of reactors

    International Nuclear Information System (INIS)

    Tamura, Takaaki.

    1984-01-01

    Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

  18. Physical model and calculation code for fuel coolant interactions

    International Nuclear Information System (INIS)

    Goldammer, H.; Kottowski, H.

    1976-01-01

    A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    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.

  20. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  1. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  2. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

  3. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

  4. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

    1994-01-01

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

  5. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  6. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  7. Effects of different rod spacers (helical types) on coolant crossmixing

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

    1981-11-01

    The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

  8. The 1994 loss of coolant incident at Pickering NGS

    Energy Technology Data Exchange (ETDEWEB)

    Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

  9. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  10. Knock-limited performance of several internal coolants

    Science.gov (United States)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  11. Impedance calculations for power cables to primary coolant pump motors

    International Nuclear Information System (INIS)

    Hegerhorst, K.B.

    1977-01-01

    The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

  12. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  13. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  14. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  15. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  16. Advanced STEM/EDX investigation on an oxide scale thermally grown on a high-chromium iron–nickel alloy under very low oxygen partial pressure

    International Nuclear Information System (INIS)

    Latu-Romain, L.; Madi, Y.; Mathieu, S.; Robaut, F.; Petit, J.-P.; Wouters, Y.

    2015-01-01

    Highlights: • A scale grown on a high-chromium iron–nickel alloy under low oxygen partial pressure was studied. • STEM-EDX maps at high resolution on a transversal thin lamella have been conducted. • The real complexity of the oxide layer has been highlighted. • These results explain the elevated number of semiconducting contributions. - Abstract: A thermal oxide scale has been grown on a high-chromium iron-nickel alloy under very low oxygen partial pressure (1050 °C, 10"−"1"0 Pa). In this paper, a special attention has been paid to morphological and chemical characterizations of the scale by scanning transmission electron microscopy and energy dispersive X-ray analysis at high resolution on a cross-section thin lamella beforehand prepared by using a combined focused ion beam/scanning electron microscope instrument. The complexity of the oxide layer is highlighted, and the correlation between the present results and the ones of a photoelectrochemical study is discussed.

  17. The Culture Repopulation Ability (CRA) Assay and Incubation in Low Oxygen to Test Antileukemic Drugs on Imatinib-Resistant CML Stem-Like Cells.

    Science.gov (United States)

    Cheloni, Giulia; Tanturli, Michele

    2016-01-01

    Chronic myeloid leukemia (CML) is a stem cell-driven disorder caused by the BCR/Abl oncoprotein, a constitutively active tyrosine kinase (TK). Chronic-phase CML patients are treated with impressive efficacy with TK inhibitors (TKi) such as imatinib mesylate (IM). However, rather than definitively curing CML, TKi induces a state of minimal residual disease, due to the persistence of leukemia stem cells (LSC) which are insensitive to this class of drugs. LSC persistence may be due to different reasons, including the suppression of BCR/Abl oncoprotein. It has been shown that this suppression follows incubation in low oxygen under appropriate culture conditions and incubation times.Here we describe the culture repopulation ability (CRA) assay, a non-clonogenic assay capable - together with incubation in low oxygen - to reveal in vitro stem cells endowed with marrow repopulation ability (MRA) in vivo. The CRA assay can be used, before moving to animal tests, as a simple and reliable method for the prescreening of drugs potentially active on CML and other leukemias with respect to their activity on the more immature leukemia cell subsets.

  18. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

  19. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  20. Nanofluid as coolant for grinding process: An overview

    Science.gov (United States)

    Kananathan, J.; Samykano, M.; Sudhakar, K.; Subramaniam, S. R.; Selavamani, S. K.; Manoj Kumar, Nallapaneni; Keng, Ngui Wai; Kadirgama, K.; Hamzah, W. A. W.; Harun, W. S. W.

    2018-04-01

    This paper reviews the recent progress and applications of nanoparticles in lubricants as a coolant (cutting fluid) for grinding process. The role of grinding machining in manufacturing and the importance of lubrication fluids during material removal are discussed. In grinding process, coolants are used to improve the surface finish, wheel wear, flush the chips and to reduce the work-piece thermal deformation. The conventional cooling technique, i.e., flood cooling delivers a large amount of fluid and mist which hazardous to the environment and humans. Industries are actively looking for possible ways to reduce the volume of coolants used in metal removing operations due to the economical and ecological impacts. Thus as an alternative, an advanced cooling technique known as Minimum Quantity Lubrication (MQL) has been introduced to the enhance the surface finish, minimize the cost, to reduce the environmental impacts and to reduce the metal cutting fluid consumptions. Nanofluid is a new-fangled class of fluids engineered by dispersing nanometre-size solid particles into base fluids such as water, lubrication oils to further improve the properties of the lubricant or coolant. In addition to advanced cooling technique review, this paper also reviews the application of various nanoparticles and their performance in grinding operations. The performance of nanoparticles related to the cutting forces, surface finish, tool wear, and temperature at the cutting zone are briefly reviewed. The study reveals that the excellent properties of the nanofluid can be beneficial in cooling and lubricating application in the manufacturing process.

  1. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  2. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  3. Coolant controls of a PEM fuel cell system

    Science.gov (United States)

    Ahn, Jong-Woo; Choe, Song-Yul

    When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

  4. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  5. Responses to Small Break Loss of Coolant Accidents for SMART

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

    2004-01-01

    The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

  6. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  7. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  8. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    International Nuclear Information System (INIS)

    Gaudard, G.; Gilles, B.; Mesnage, F.; Cattant, F.

    2002-01-01

    In the framework of a radiation exposure management program entitled >, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on 110m Ag, 122 Sb, 124 Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  9. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    Science.gov (United States)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  10. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  11. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  12. A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

  13. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  14. Ultrastructure and molecular phylogenetic position of a novel phagotrophic stramenopile from low oxygen environments: Rictus lutensis gen. et sp. nov. (Bicosoecida, incertae sedis).

    Science.gov (United States)

    Yubuki, Naoji; Leander, Brian S; Silberman, Jeffrey D

    2010-04-01

    A novel free free-living phagotrophic flagellate, Rictus lutensis gen. et sp. nov., with two heterodynamic flagella, a permanent cytostome and a cytopharynx was isolated from muddy, low oxygen coastal sediments in Cape Cod, MA, USA. We cultivated and characterized this flagellate with transmission electron microscopy, scanning electron microscopy and molecular phylogenetic analyses inferred from small subunit (SSU) rDNA sequences. These data demonstrated that this organism has the key ultrastructural characters of the Bicosoecida, including similar transitional zones and a similar overall flagellar apparatus consisting of an x fiber and an L-shape microtubular root 2 involved in food capture. Although the molecular phylogenetic analyses were concordant with the ultrastructural data in placing R. lutensis with the bicosoecid clade, the internal position of this relatively divergent sequence within the clade was not resolved. Therefore, we interpret R. lutensis gen. et sp. nov. as a novel bicosoecid incertae sedis. Copyright 2009 Elsevier GmbH. All rights reserved.

  15. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  16. Vision system for precision alignment of coolant channels

    International Nuclear Information System (INIS)

    Kar, S.; Rao, Y.V.; Valli Kumar; Joshi, D.G.; Chadda, V.K.; Nigam, R.K.; Kayal, J.N.; Panwar, S.; Sinha, R.K.

    1997-01-01

    This paper describes a vision system which has been developed for precision alignment of Coolant Channel Replacement Machine (CCRM) with respect to the front face of the coolant channel under repair/replacement. It has provisions for automatic as well as semi-automatic alignment. A special lighting scheme has been developed for providing illumination to the front face of the channel opening. This facilitates automatic segmentation of the digitized image. The segmented image is analysed to obtain the centre of the front face of the channel opening and thus the extent of misalignment i.e. offset of the camera with respect to the front face of the channel opening. The offset information is then communicated to the PLC to generate an output signal to drive the DC servo motors for precise positioning of the co-ordinate table. 2 refs., 5 figs

  17. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  18. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  19. Rapid thermal transient in a reactor coolant channel

    International Nuclear Information System (INIS)

    Cherubini, A.

    1986-01-01

    This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

  20. Coolant Chemistry Control: Oxygen Mass Transport in Lead Bismuth Eutectic

    International Nuclear Information System (INIS)

    Weisenburger, A.; Mueller, G.; Bruzzese, C.; Glass, A.

    2015-01-01

    In lead-bismuth cooled transmutation systems, oxygen, dissolved in the coolant at defined quantities, is required for stable long-term operation by assuring the formation of protective oxide scales on structural steel surfaces. Extracted oxygen must be permanently delivered to the system and distributed in the entire core. Therefore, coolant chemistry control involves detailed knowledge on oxygen mass transport. Beside the different flow regimes a core might have stagnant areas at which oxygen delivery can only be realised by diffusion. The difference between oxygen transport in flow paths and in stagnant zones is one of the targets of such experiments. To investigate oxygen mass transport in flowing and stagnant conditions, a dedicated facility was designed based on computational fluid dynamics (CFD). CFD also was applied to define the position of oxygen sensors and ultrasonic Doppler velocimetry transducers for flow measurements. This contribution will present the test facility, design relevant CFD calculations and results of first tests performed. (authors)

  1. Health physics aspects of processing EBR-I coolant

    International Nuclear Information System (INIS)

    Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

    1998-01-01

    The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed

  2. Qualification test of a main coolant pump for SMART pilot

    International Nuclear Information System (INIS)

    Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

    2006-01-01

    SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

  3. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  4. Device for preventing leakage of coolant in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kobayashi, Yukio; Sekiguchi, Mamoru; Yoshida, Hideo.

    1975-01-01

    Object: To prevent leakage of coolant from between lower tie plate and channel box without causing deformation of the channel box and also without the possibility of disturbing the installation and removal of the box by the provision of a thin plate provided with leakage holes for the lower tie plate. Structure: Static water pressure within the lower tie plate is adapted to act upon the bear side of a flat plate for leakage prevention through leakage holes formed in the tie plate, thus urging the flat plate against the channel box inner surface. Meanwhile, static water pressure having been led through the leakage holes in the flat plate is adapted to press the flat plate in the vertical direction, thus urging the flat plate against the channel box inner surface and thereby preventing leakage of the coolant through a gap between the channel box and lower tie plate. (Yoshino, Y.)

  5. Sound velocity in the coolant of boiling nuclear reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

  6. Calculation of coolant temperature sensitivity related to thermohydraulic parameters

    International Nuclear Information System (INIS)

    Silva, F.C. da; Andrade Lima, F.R. de

    1985-01-01

    It is verified the viability to apply the generalized Perturbation Theory (GPT) in the calculation of sensitivity for thermal-hydraulic problems. It was developed the TEMPERA code in FORTRAN-IV to transient calculations in the axial temperature distribution in a channel of PWR reactor and the associated importance function, as well as effects of variations of thermalhydraulic parameters in the coolant temperature. The results are compared with one which were obtained by direct calculation. (M.C.K.) [pt

  7. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  8. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  9. Structural integrity analysis of reactor coolant pump flywheel(I)

    International Nuclear Information System (INIS)

    Kim, Young Jin

    1986-01-01

    A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

  10. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  11. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  12. Design of Reactor Coolant Pump Seal Online Monitoring System

    International Nuclear Information System (INIS)

    Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

    2008-01-01

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

  13. Detection of coolant void in lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

    2015-01-01

    Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

  14. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  15. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  16. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  17. On-line real time gamma analysis of primary coolant

    International Nuclear Information System (INIS)

    Kalechstein, W.; Kupca, S.; Lipsett, J.J.

    1985-10-01

    The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

  18. Labelling Of Coolant Flow Anomaly Using Fractal Structure

    International Nuclear Information System (INIS)

    Djainal, Djen Djen

    1996-01-01

    This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

  19. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  20. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  1. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  2. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  3. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  4. Feasibility study on the type of KALIMER coolant circulation pump

    International Nuclear Information System (INIS)

    Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

    1997-07-01

    The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

  5. A open-quotes zero wasteclose quotes coolant management strategy

    International Nuclear Information System (INIS)

    Kennicott, M.A.

    1994-01-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant's Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant's liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study's goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a open-quotes zero wasteclose quotes machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs

  6. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  7. Analysis of actual status of works on technology of heavy liquid metal coolants

    International Nuclear Information System (INIS)

    Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

    2014-01-01

    Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

  8. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  9. Membrane technology for treating of waste nanofluids coolant: A review

    Science.gov (United States)

    Mohruni, Amrifan Saladin; Yuliwati, Erna; Sharif, Safian; Ismail, Ahmad Fauzi

    2017-09-01

    The treatment of cutting fluids wastes concerns a big number of industries, especially from the machining operations to foster environmental sustainability. Discharging cutting fluids, waste through separation technique could protect the environment and also human health in general. Several methods for the separation emulsified oils or oily wastewater have been proposed as three common methods, namely chemical, physicochemical and mechanical and membrane technology application. Membranes are used into separate and concentrate the pollutants in oily wastewater through its perm-selectivity. Meanwhile, the desire to compensate for the shortcomings of the cutting fluid media in a metal cutting operation led to introduce the using of nanofluids (NFs) in the minimum quantity lubricant (MQL) technique. NFs are prepared based on nanofluids technology by dispersing nanoparticles (NPs) in liquids. These fluids have potentially played to enhance the performance of traditional heat transfer fluids. Few researchers have studied investigation of the physical-chemical, thermo-physical and heat transfer characteristics of NFs for heat transfer applications. The use of minimum quantity lubrication (MQL) technique by NFs application is developed in many metal cutting operations. MQL did not only serve as a better alternative to flood cooling during machining operation and also increases better-finished surface, reduces impact loads on the environment and fosters environmental sustainability. Waste coolant filtration from cutting tools using membrane was treated by the pretreated process, coagulation technique and membrane filtration. Nanomaterials are also applied to modify the membrane structure and morphology. Polyvinylidene fluoride (PVDF) is the better choice in coolant wastewater treatment due to its hydrophobicity. Using of polyamide nanofiltration membranes BM-20D and UF-PS-100-100, 000, it resulted in the increase of permeability of waste coolant filtration. Titanium dioxide

  10. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  11. Numerical FEM Analyses of primary coolant system at NPP Temelin

    International Nuclear Information System (INIS)

    Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

    2003-01-01

    The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

  12. Fuel-coolant interaction-phenomena under prompt burst conditions

    International Nuclear Information System (INIS)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

  13. Modelling transient energy release from molten fuel coolant interaction debris

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-05-01

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

  14. In-operation diagnostic system for reactor coolant pump

    International Nuclear Information System (INIS)

    Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

    1996-01-01

    A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

  15. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

  16. Graphite beds for coolant filtration at high temperature

    International Nuclear Information System (INIS)

    Heathcock, R.E.; Lacy, C.S.

    1978-01-01

    High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

  17. Effects contributing to positive coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Garland, W.J.; Milgram, M.S.

    1995-01-01

    The lattice cell code WIMS-AECL (Ref. 3) is used to model a typical CANDU lattice cell, using nominal geometric bucklings, the PIJ option, and 69-group Winfrith library. The effect of cell voiding is modeled as a 100% instantaneous removal of coolant from the lattice. This is conservative because of the neglect of time dependence and partial core voiding, considered more plausible in CANDU. Results are grouped into three spectral groups: fast neutrons (0.821 to 10 MeV), epithermal neutrons (0.625 eV to 0.821 MeV), and thermal neutrons (<0.625 eV)

  18. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  19. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  20. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  1. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  2. Comparison of thermohydraulic characteristics in the use of various coolants

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

    2000-11-01

    Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet

  3. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials

  4. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

  5. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  6. Some observations on simulated molten debris-coolant layer dynamics

    International Nuclear Information System (INIS)

    Greene, G.A.; Klein, J.; Klages, J.; Schwarz, E.; Sanborn, Y.

    1983-04-01

    Experiments are being performed to investigate high temperature liquid-liquid film boiling between a pool of liquid metal and an overlying coolant pool of R-11 or water. Film boiling has been observed to be stable for R-11; however, considerable liquid-liquid contact has been observed with water well beyond the minimum film boiling temperature. Unstable liquid-liquid film boiling of water has been observed to escalate into dispersive, non-energetic vapor explosions when the interface contact temperature exceeded the spontaneous nucleation temperature. Other parametric trends in the data are discussed

  7. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  8. Void fraction calculation in a channel containing boiling coolant

    International Nuclear Information System (INIS)

    Norelli, F.

    1978-01-01

    The problem of void fraction calculation was studied for a channel containing boiling coolant, when a slip ratio correlation is used. Use of fitting (e.g. polinomial or rational algebraic) for slip ratio correlation and the characteristic method are proposed in this work. In this way we are reduced to some elementary quadrature problem. Another problem discussed in the present work concerns what we must consider as ''initial condition'' in any initial value problem, in order to take into account different error distributions in steady state and in successive time-dependent calculations

  9. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  10. Reactor coolant pump shaft seal stability during station blackout

    International Nuclear Information System (INIS)

    Rhodes, D.B.; Hill, R.C.; Wensel, R.G.

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries

  11. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  12. Evaluation of conservatism in analysis of fuel-coolant interaction

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Haas, P.M.; Allen, C.L.

    Using the ANL parametric model developed by Cho e.a. the following mechanisms and parameters involved in fuel-coolant interaction were examined: coherence of fuel-sodium mixing; two-phase heat transfer; sodium-to-fuel mass ratio; fuel particle size; heat transfer to plenum and core cladding; constraint geometry. Both overpower and loss-of-flow transients were studied. Main attention is given to the maximum mechanical work to be expected. As a general conclusion, it can be stated that more realistic models will result in a reduction of the estimated mechanical work

  13. Newly developed EMF cell with zirconia solid electrolyte for measurement of low oxygen potentials in liquid Cu-Cr and Cu-Zr alloys

    Directory of Open Access Journals (Sweden)

    Katayama I.

    2012-01-01

    Full Text Available In order to measure the very low oxygen potential by use of stabilized zirconia solid electrolyte emf method, a new cell construction was devised. The idea was based on Janke but a zirconia rod was used instead of the zirconia crucible which contacts liquid alloy electrode. The cell was used for determination of the oxygen potentials in liquid dilute Cu-Cr and Cu-Zr alloys. The reference electrode was Cr,Cr2O3. Emf measurements were performed in the temperature range of 1400-1580K and composition range of 0.198-3.10at%Cr-Cu alloys, and 1380-1465K, 0.085-0.761at%Zr-Cu alloys. The composition of liquid alloys were determined by picking up from the liquid alloys and ICP analysis. By use of the newly devised cell construction in this study, stable emf values were obtained at each temperature and alloy composition. Emf values were corrected by using the parameter for electronic contribution of the YSZ. Activity of Cr obeys Henry’s law and activity coefficient at infinitely dilute alloys of Cr in Cu-Cr alloys are: lng0 Cr =(3.80 at 1423K, (3.57 at 1473K, (3.38 at 1523K and (3.20 at 1573K. At 1423 K activity coefficient of Zr at infinitely diluted alloy is lnγo Zr = -4.0.

  14. Exposure to elevated pCO2 does not exacerbate reproductive suppression of Aurelia aurita jellyfish polyps in low oxygen environments

    KAUST Repository

    Treible, LM

    2017-08-15

    Eutrophication-induced hypoxia is one of the primary anthropogenic threats to coastal ecosystems. Under hypoxic conditions, a deficit of O2 and a surplus of CO2 will concurrently decrease pH, yet studies of hypoxia have seldom considered the potential interactions with elevated pCO2 (reduced pH). Previous studies on gelatinous organisms concluded that they are fairly robust to low oxygen and reduced pH conditions individually, yet the combination of stressors has only been examined for ephyrae. The goals of this study were to determine the individual and interactive effects of hypoxia and elevated pCO2 on the asexual reproduction and aerobic respiration rates of polyps of the scyphozoan Aurelia aurita during a manipulative experiment that ran for 36 d. pCO2 and pO2 were varied on a diel basis to closely mimic the diel conditions observed in the field. Exposure to low dissolved oxygen (DO) reduced asexual budding of polyps by ~50% relative to control conditions. Under hypoxic conditions, rates of respiration were elevated during an initial acclimation period (until Day 8), but respiration rates did not differ between DO levels under prolonged exposure. There was no significant effect of increased pCO2 on either asexual reproduction or aerobic respiration, suggesting that elevated pCO2 (reduced pH) did not exacerbate the negative reproductive effects of hypoxia on A. aurita polyps.

  15. Exposure to elevated pCO2 does not exacerbate reproductive suppression of Aurelia aurita jellyfish polyps in low oxygen environments

    KAUST Repository

    Treible, LM; Pitt, KA; Klein, SG; Condon, RH

    2017-01-01

    Eutrophication-induced hypoxia is one of the primary anthropogenic threats to coastal ecosystems. Under hypoxic conditions, a deficit of O2 and a surplus of CO2 will concurrently decrease pH, yet studies of hypoxia have seldom considered the potential interactions with elevated pCO2 (reduced pH). Previous studies on gelatinous organisms concluded that they are fairly robust to low oxygen and reduced pH conditions individually, yet the combination of stressors has only been examined for ephyrae. The goals of this study were to determine the individual and interactive effects of hypoxia and elevated pCO2 on the asexual reproduction and aerobic respiration rates of polyps of the scyphozoan Aurelia aurita during a manipulative experiment that ran for 36 d. pCO2 and pO2 were varied on a diel basis to closely mimic the diel conditions observed in the field. Exposure to low dissolved oxygen (DO) reduced asexual budding of polyps by ~50% relative to control conditions. Under hypoxic conditions, rates of respiration were elevated during an initial acclimation period (until Day 8), but respiration rates did not differ between DO levels under prolonged exposure. There was no significant effect of increased pCO2 on either asexual reproduction or aerobic respiration, suggesting that elevated pCO2 (reduced pH) did not exacerbate the negative reproductive effects of hypoxia on A. aurita polyps.

  16. Fabrication of lead-free piezoelectric Li2CO3-added (Ba,Ca)(Ti,Sn)O3 ceramics under controlled low oxygen partial pressure and their properties

    Science.gov (United States)

    Noritake, Kouta; Sakamoto, Wataru; Yuitoo, Isamu; Takeuchi, Teruaki; Hayashi, Koichiro; Yogo, Toshinobu

    2018-02-01

    Reduction-resistant lead-free (Ba,Ca)(Ti,Sn)O3 piezoceramics with high piezoelectric constants were fabricated by optimizing the amount of Li2CO3 added. Oxygen partial pressure was controlled during the sintering of (Ba,Ca)(Ti,Sn)O3 ceramics in a reducing atmosphere using H2-CO2 gas. Enhanced grain growth and a high-polarization state after poling treatment were achieved by adding Li2CO3. Optimizing the amount of Li2CO3 added to (Ba0.95Ca0.05)(Ti0.95Sn0.05)O3 ceramics sintered under a low oxygen partial pressure resulted in improved piezoelectric properties while maintaining the high sintered density. The prepared Li2CO3-added ceramic samples had homogeneous microstructures with a uniform dispersion of each major constituent element. However, the residual Li content in the 3 mol % Li2CO3-added (Ba0.95Ca0.05)(Ti0.95Sn0.05)O3 ceramics after sintering was less than 0.3 mol %. Sintered bodies of this ceramic prepared in a CO2 (1.5%)-H2 (0.3%)/Ar reducing atmosphere (PO2 = 10-8 atm at 1350 °C), exhibited sufficient electrical resistivity and a piezoelectric constant (d 33) exceeding 500 pC/N. The piezoelectric properties of this nonreducible ceramic were comparable or superior to those of the same ceramic sintered in air.

  17. Effect of low-oxygen-concentration layer on iron gettering capability of carbon-cluster ion-implanted Si wafer for CMOS image sensors

    Science.gov (United States)

    Onaka-Masada, Ayumi; Nakai, Toshiro; Okuyama, Ryosuke; Okuda, Hidehiko; Kadono, Takeshi; Hirose, Ryo; Koga, Yoshihiro; Kurita, Kazunari; Sueoka, Koji

    2018-02-01

    The effect of oxygen (O) concentration on the Fe gettering capability in a carbon-cluster (C3H5) ion-implanted region was investigated by comparing a Czochralski (CZ)-grown silicon substrate and an epitaxial growth layer. A high Fe gettering efficiency in a carbon-cluster ion-implanted epitaxial growth layer, which has a low oxygen region, was observed by deep-level transient spectroscopy (DLTS) and secondary ion mass spectroscopy (SIMS). It was demonstrated that the amount of gettered Fe in the epitaxial growth layer is approximately two times higher than that in the CZ-grown silicon substrate. Furthermore, by measuring the cathodeluminescence, the number of intrinsic point defects induced by carbon-cluster ion implantation was found to differ between the CZ-grown silicon substrate and the epitaxial growth layer. It is suggested that Fe gettering by carbon-cluster ion implantation comes through point defect clusters, and that O in the carbon-cluster ion-implanted region affects the formation of gettering sinks for Fe.

  18. Design technology development of the main coolant pump for an integral reactor

    International Nuclear Information System (INIS)

    Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

    2004-01-01

    All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

  19. Preliminary design of reactor coolant pump canned motor for AC600

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

    The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

  20. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  1. Determination of temperature distributions in fast reactor core coolants

    International Nuclear Information System (INIS)

    Tillman, M.

    1975-04-01

    An analytical method of determination of a temperature distribution in the coolant medium in a fuel assembly of a liquid-metal-fast-breeder-reactor (LMFBR) is presented. The temperature field obtained is applied for a constant velocity (slug flow) fluid flowing, parallel to the fuel pins of a square and hexagonal array assembly. The coolant subchannels contain irregular boundaries. The geometry of the channel due to the rod adjacent to the wall (edge rod) differs from the geometry of the other channels. The governing energy equation is solved analytically, assuming series solutions for the Poisson and diffusion equations, and the total solution is superposed by the two. The boundary conditions are specified by symmetry considerations, assembly wall insulation and a continuity of the temperature field and heat fluxes. The initial condition is arbitrary. The method satisfies the boundary conditions on the irregular boundaries and the initial condition by a least squares technique. Computed results are presented for various geometrical forms, with ratio of rod pitch-to-diameter typical for LMFBR cores. These results are applicable for various fast-reactors, and thus the influence of the transient solution (which solves the diffusion equation) on the total depends on the core parameters. (author)

  2. Contact condensation effects in the main coolant pipe

    International Nuclear Information System (INIS)

    Haefner, W.; Fischer, K.

    1990-01-01

    Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

  3. Multi-objective optimization of the reactor coolant system

    International Nuclear Information System (INIS)

    Chen Lei; Yan Changqi; Wang Jianjun

    2014-01-01

    Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

  4. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  5. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  6. Reactor coolant purification system circulation pumps (CUW pumps)

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

    Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

  7. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  8. Measuring device for the coolant flowrate in a reactor core

    International Nuclear Information System (INIS)

    Sawa, Toshihiko.

    1983-01-01

    Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

  9. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    International Nuclear Information System (INIS)

    Vukovic, J.; Besirevic, A.; Boljat, Z.

    2016-01-01

    After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

  10. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    Debbage, A.G.; Driver, M.; Waller, P.R.

    1964-01-01

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  11. Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

    International Nuclear Information System (INIS)

    Albert K Papovyants; Yuri I Orlov; Pyotr N Martynov; Yuri D Boltoev

    2005-01-01

    Full text of publication follows: To optimize the design of filters for cleaning heavy liquid metal coolant (HLMC) from suspended impurities and choose appropriate filter material, the contribution is considered of different mechanisms of delivery and retention of these impurities from the coolant flow, which is governed by its specificity as a thermodynamically instable disperse system to a large extent. It is shown that the buildup of deposits in the filter is favored by the hydrodynamic regime with minimum filtration rates being due to the predominance in the suspension of the fine-dispersed solid phase (oxides Fe 3 O 4 , Cr 2 O 3 and so on). With concentrating the last mentioned phase in filter material pores or stagnant zones, coagulation structuration is possible, which is accompanied by sharp local increase in the viscosity and strength of the solid phase medium being built from liquid metal, i.e. slag sedimentary deposits. In rather extended pores, disintegration of such structures is possible, which is accompanied by sedimentation of large particles produced due to sticking together at coagulation. The analytical solution of the problem of particle sedimentation due to diffusion indicated that in the case under consideration, this mechanism takes place for particles less than ∼ 0,05 μm in size, which is specified by the fact that the time of their delivery to the filter material surface is longer than that of the coolant being in the filter. The London-Van-der-Waals molecular forces play a crucial role in the stage of retention of a separate particle. The constant of the molecular interaction between a spherical particle and the flat surface has been estimated for the chosen value of the gap between the contacting bodies, being dependent on the wetting angle. The sufficient condition for d p -diameter particle capture by the adhesion force field (with a gap of H ≅ 30 nm) is that it be brought by the appropriate forces at a distance from the wall equal

  12. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

  13. Multi-state reliability for coolant pump based on dependent competitive failure model

    International Nuclear Information System (INIS)

    Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

    2013-01-01

    By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

  14. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  15. Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

  16. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1997-01-01

    This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

  17. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  18. Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

    International Nuclear Information System (INIS)

    Monta, Kazuo.

    1974-01-01

    Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

  19. Effects of in vitro low oxygen tension preconditioning of adipose stromal cells on their in vivo chondrogenic potential: application in cartilage tissue repair.

    Directory of Open Access Journals (Sweden)

    Sophie Portron

    Full Text Available PURPOSE: Multipotent stromal cell (MSC-based regenerative strategy has shown promise for the repair of cartilage, an avascular tissue in which cells experience hypoxia. Hypoxia is known to promote the early chondrogenic differentiation of MSC. The aim of our study was therefore to determine whether low oxygen tension could be used to enhance the regenerative potential of MSC for cartilage repair. METHODS: MSC from rabbit or human adipose stromal cells (ASC were preconditioned in vitro in control or chondrogenic (ITS and TGF-β medium and in 21 or 5% O2. Chondrogenic commitment was monitored by measuring COL2A1 and ACAN expression (real-time PCR. Preconditioned rabbit and human ASC were then incorporated into an Si-HPMC hydrogel and injected (i into rabbit articular cartilage defects for 18 weeks or (ii subcutaneously into nude mice for five weeks. The newly formed tissue was qualitatively and quantitatively evaluated by cartilage-specific immunohistological staining and scoring. The phenotype of ASC cultured in a monolayer or within Si-HPMC in control or chondrogenic medium and in 21 or 5% O2 was finally evaluated using real-time PCR. RESULTS/CONCLUSIONS: 5% O2 increased the in vitro expression of chondrogenic markers in ASC cultured in induction medium. Cells implanted within Si-HPMC hydrogel and preconditioned in chondrogenic medium formed a cartilaginous tissue, regardless of the level of oxygen. In addition, the 3D in vitro culture of ASC within Si-HPMC hydrogel was found to reinforce the pro-chondrogenic effects of the induction medium and 5% O2. These data together indicate that although 5% O2 enhances the in vitro chondrogenic differentiation of ASC, it does not enhance their in vivo chondrogenesis. These results also highlight the in vivo chondrogenic potential of ASC and their potential value in cartilage repair.

  20. Laboratory Study on Prevention of CaO-Containing ASTM "D-Type" Inclusions in Al-Deoxidized Low-Oxygen Steel Melts During Basic Slag Refining

    Science.gov (United States)

    Jiang, Min; Wang, Xin-Hua; Yang, Die; Lei, Shao-Long; Wang, Kun-Peng

    2015-12-01

    Present work was attempted to explore the possibility of preventing CaO-containing inclusions in Al-deoxidized low-oxygen special steel during basic slag refining, which were known as ASTM D-type inclusions. Based on the analysis on formation thermodynamics of CaO-containing inclusions, a series of laboratory experiments were designed and carried out in a vacuum induction furnace. During the experiments, slag/steel reaction equilibrium was intentionally suppressed with the aim to decrease the CaO contents in inclusions, which is different from ordinary concept that slag/steel reaction should be promoted for better control of inclusions. The obtained results showed that high cleanliness of steel was obtained in all the steel melts, with total oxygen contents varied between 0.0003 and 0.0010 pct. Simultaneously, formation of CaO-containing inclusions was successfully prohibited, and all the formed oxide inclusions were MgO-Al2O3 or/and Al2O3 in very small sizes of about 1 to 3 μm. And 90 pct to nearly 98 pct of them were wrapped by relative thicker MnS outer surface layers to produce dual-phased "(MgO-Al2O3) + MnS" or "Al2O3 + MnS" complex inclusions. Because of much better ductility of MnS, certain deformability of these complex inclusions can be expected which is helpful to improve fatigue resistance property of steel. Only very limited number of singular MnS inclusions were with sizes larger than 13 μm, which were formed during solidification because of. In the end, formation of oxide inclusions in steel was qualitatively evaluated and discussed.

  1. Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

  2. Zero waste machine coolant management strategy at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-01-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

  3. New method for NPP sodium coolant pipeline austenization

    International Nuclear Information System (INIS)

    Malashonok, V.A.; Rotshtejn, A.V.; Gotshalk, A.L.; Miryushchenko, E.F.

    1980-01-01

    Heat treatment technology is considered for pipelines intended for the NPP cooling systems employing sodium coolant. Various techniques are discussed which are used for protecting the pipeline internal surfaces against oxidation in the process of heat treatment. It is noted that the austenite formation of welded joints of steel 12Kh18N9 and steel Kh16N11M3 at temperatures of 1050 and 1100 deg C releases welding-induced stresses and reduces a possibility of local damages. Evacuation down to 1 mm Hg appears to be the most rational protective technique. The considered procedure of the pipeline heat treatment has been utilized for mounting the equipment of the BN-600 reactor at the Beloyarskaya NPP. The economic gain resulting from the use of the procedure, owing to decrease in argon consumption and reduction of labour input, makes up 150 000 roubles

  4. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  5. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  6. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    International Nuclear Information System (INIS)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs

  7. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  8. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  9. Fatigue check of nuclear safety class 1 reactor coolant pipe

    International Nuclear Information System (INIS)

    Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

    2015-01-01

    Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

  10. Fuel -coolant interactions in LWRs and LMFBRs: relationships and distinctions

    Energy Technology Data Exchange (ETDEWEB)

    Duffey, R B; Lellouche, G S [Nuclear Safety and Analysis Department, Electric Power Research Institute, Palo Alto, CA (United States)

    1979-10-15

    The question of fuel-coolant interaction and of potential vapor explosion is raised here. lt is the contention of the authors that there is in fact no need to study this question vis a vis Light Water Reactors (LWR) except from an academic point of view since it does not impact on safety considerations. As for LMFBRs, the design basis whole core accidents for LWRs are derived from the fundamental concern of maintaining core geometry to provide for convective cooling. However, the important distinction is that the core is in its most reactive configuration, and core and fuel rearrangement is therefore not of such concern. The author's thesis is that even if the probability of steam explosion following core melt were two orders of magnitude greater than currently assumed (10{sup -2}) the total LWR risk would increase only by a factor of 2-6 for BWRs and less a factor of 10 for PWRs

  11. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

    International Nuclear Information System (INIS)

    Bayless, P.D.; Dobbe, C.A.

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  12. Fuel-coolant interactions in a jet contact mode

    International Nuclear Information System (INIS)

    Konishi, K.; Isozaki, M.; Imahori, S.; Kondo, S.; Furutani, A.; Brear, D.J.

    1994-01-01

    Molten fuel-coolant interactions in a jet contact mode was studied with respect to the safety of liquid-metal-cooled fast reactors (LMFRs). From a series of molten Wood's metal (melting point: 79 deg. C, density: -8400 kg/m 3 ) jet-water interaction experiments, several distinct modes of interaction behaviors were observed for various combinations of initial temperature conditions of the two fluids. A semi-empirical model for a minimum film boiling temperature criterion was developed and used to reasonably explain the different interaction modes. It was concluded that energetic jet-water interactions are only possible under relatively narrow initial thermal conditions. Preliminary extrapolation of the present results in an oxide fuel-sodium system suggests that mild interactions with short breakup length and coolable debris formation should be most likely in LMFRs. (author)

  13. A new inexpensive electrochemical meter for oxygen in sodium coolant

    International Nuclear Information System (INIS)

    Periaswami, G.; Rajan Babu, S.S.; Mathews, C.K.

    1987-01-01

    This report describes the development of an inexpensive oxygen meter for sodium coolant and gives the results of the test experiments. Calcia stabilized zirconia has been found to have necessary domain boundary characteristics at low temperatures for use as oxygen sensor in liquid sodium system. It is possible to obtain acceptable sensor cell resistance at temperatures as low as 230 C if K, K 2 O or Na, Na 2 O is used as reference electrode. The performance of these cells has been tested in bench top sodium loops over long periods. Their performance in terms of cell-out put variation with change in oxygen concentration in sodium has been found to be satisfactory. They also have sufficiently long life times since the kinetics of sodium attack on the electrolyte is slow at low temperatures. (author). 17 refs., 6 figs

  14. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

    DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-10-02

    The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

  15. On-line monitoring of main coolant pump seals

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

    1984-06-01

    The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

  16. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

  17. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  18. Reactor coolant pump seal response to loss of cooling

    International Nuclear Information System (INIS)

    Graham, T.; Metcalfe, R.; Burchett, P.

    2000-01-01

    This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

  19. Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

    International Nuclear Information System (INIS)

    Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

    1998-07-01

    At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

  20. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D P; Lukas, M; Lynch, B K [Spectro Incorporated, Littleton, MA (United States)

    1998-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  1. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  2. Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

    2000-01-01

    The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

  3. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

  4. Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

    Science.gov (United States)

    Salamon, V.; Senthil kumar, D.; Thirumalini, S.

    2017-08-01

    The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

  5. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  6. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  7. Operation diagnostics of the reactor coolant pumps in the Jaslovske Bohunice nuclear power plant, CSSR

    International Nuclear Information System (INIS)

    Bahna, J.; Jaros, I.; Oksa, G.

    1990-01-01

    The state of the art of the materials basis, the diagnostics methods used, organization of data collection and processing, and some results of routine and specific investigations concerned with diagnosis of the reactor coolant pump in the Jaslovske Bohunice NPP V-1 are presented. Some information is given about the reactor coolant pump monitor developed in the VUJE. (author)

  8. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  9. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  10. A comparative neutronic analysis of 150MWe TRU burner according to the coolant alteration

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic analysis has been conducted for the small TRU burner according to their coolant material. The use of Pb-Bi coolant gave a low burnup reactivity swing and negative or less positive coolant void coefficient with harder neutron spectrum. By a lower burnup reactivity swing and higher conversion ratio of Pb-Bi cooled core, the total amount of TRU consumption was found to be small compared with Na cooled core despite of the higher MA consumption ratio of Pb-Bi cooled core. However, Pb-Bi cooled reactor have a lager margin in the coolant void coefficient, so that a variable MA composition can be loaded in the core. Accordingly, even though the Pb-Bi cooled TRU burner has not effectiveness on TRU burning in the same geometry and material condition, a flexible MA loading is envisaged to result in 10 times larger MA burning amount, still preserving a low coolant void worth

  11. Radioactive corrosion products in circuit of fast reactor loop with dissociating coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.

    1982-01-01

    The results of experimental investigation into depositions of radionuclides of corrosion origin on the surfaces of a reactor-in-pile loop facility with a dissociating coolant are presented. It is stated that the ratio of radionuclides in fixed depositions linearly decreases with decrease of the coolant temperature at the core-condenser section. The element composition of non-fixed compositions quantitatively and qualitatively differs from the composition of structural material, and it is more vividly displayed for the core-condenser section. The main mechanism of circuit contamination with radioactive corrosion products is substantiated: material corrosion in the zones of coolant phase transfer, their remove by the coolant in the core, deposition, activation and wash-out by the coolant from the core surfaces

  12. Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

    International Nuclear Information System (INIS)

    Ruger, C.J.; Luckas, W.J. Jr.

    1989-03-01

    Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

  13. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  14. Geochemistry of trace metals in shelf sediments affected by seasonal and permanent low oxygen conditions off central Chile, SE Pacific (˜36°S)

    Science.gov (United States)

    Muñoz, Praxedes; Dezileau, Laurent; Cardenas, Lissette; Sellanes, Javier; Lange, Carina B.; Inostroza, Jorge; Muratli, Jesse; Salamanca, Marco A.

    2012-02-01

    Trace metals (Cd, U, Co, Ni, Cu, Ba, Fe, Mn), total organic carbon (TOC) and C and N stable isotope signatures (δ 13C and δ 15N) were determined in short sediments cores from the inner and outer shelf off Concepción, Chile (˜36°S). The objectives were to establish the effect of environmental conditions on trace metal distributions at two shelf sites, one affected by seasonal oxygenation and the other by permanent low oxygen conditions due to the presence of the oxygen minimum zone (OMZ). We evaluate trace metals as proxies of past changes in primary productivity and the bottom water oxygen regime. Concentrations of pore water sulfides and NH4+ were also measured as indicators of the main diagenetic pathways at each site. Our results for the inner shelf (seasonal suboxia) suggest that the oxidative state of the sediments responds to seasonal pulses of organic matter and that seasonal oxygenation develops during high and low primary productivity in the water column. Here, positive fluxes (to the water column) estimated from pore water concentrations of several elements were observed (Ba, Co, Ni, Fe and Mn). The less reduced environment at this site produces authigenic enrichment of Cu associated with the formation of oxides in the oxic surface sediment layer, and the reduction of U within deeper sediment sections occur consistently with negative estimated pore water fluxes. In the outer shelf sediments (permanent suboxia, OMZ site), negative fluxes (to the sediment) were estimated for all elements, but these sediments showed authigenic enrichments only for Cd, Cu and U. The short oxygenation period during the winter season did not affect the accumulation of these metals on the shelf. The distribution of Cu, Cd and U have been preserved within the sediments and the authigenic accumulation rates estimated showed a decrease from the deep sections of the core to the surface sediments. This could be explained by a gradual decrease in the strength of the OMZ in the

  15. Growth of the obligate anaerobe Desulfovibrio vulgaris Hildenborough under continuous low oxygen concentration sparging: impact of the membrane-bound oxygen reductases.

    Science.gov (United States)

    Ramel, Fanny; Brasseur, Gael; Pieulle, Laetitia; Valette, Odile; Hirschler-Réa, Agnès; Fardeau, Marie Laure; Dolla, Alain

    2015-01-01

    Although obligate anaerobe, the sulfate-reducing bacterium Desulfovibrio vulgaris Hildenborough (DvH) exhibits high aerotolerance that involves several enzymatic systems, including two membrane-bound oxygen reductases, a bd-quinol oxidase and a cc(b/o)o3 cytochrome oxidase. Effect of constant low oxygen concentration on growth and morphology of the wild-type, single (Δbd, Δcox) and double deletion (Δcoxbd) mutant strains of the genes encoding these oxygen reductases was studied. When both wild-type and deletion mutant strains were cultured in lactate/sulfate medium under constant 0.02% O2 sparging, they were able to grow but the final biomasses and the growth yield were lower than that obtained under anaerobic conditions. At the end of the growth, lactate was not completely consumed and when conditions were then switched to anaerobic, growth resumed. Time-lapse microscopy revealed that a large majority of the cells were then able to divide (over 97%) but the time to recover a complete division event was longer for single deletion mutant Δbd than for the three other strains. Determination of the molar growth yields on lactate suggested that a part of the energy gained from lactate oxidation was derived toward cells protection/repairing against oxidative conditions rather than biosynthesis, and that this part was higher in the single deletion mutant Δbd and, to a lesser extent, Δcox strains. Our data show that when DvH encounters oxidative conditions, it is able to stop growing and to rapidly resume growing when conditions are switched to anaerobic, suggesting that it enters active dormancy sate under oxidative conditions. We propose that the pyruvate-ferredoxin oxidoreductase (PFOR) plays a central role in this phenomenon by reversibly switching from an oxidative-sensitive fully active state to an oxidative-insensitive inactive state. The oxygen reductases, and especially the bd-quinol oxidase, would have a crucial function by maintaining reducing conditions

  16. Growth of the obligate anaerobe Desulfovibrio vulgaris Hildenborough under continuous low oxygen concentration sparging: impact of the membrane-bound oxygen reductases.

    Directory of Open Access Journals (Sweden)

    Fanny Ramel

    Full Text Available Although obligate anaerobe, the sulfate-reducing bacterium Desulfovibrio vulgaris Hildenborough (DvH exhibits high aerotolerance that involves several enzymatic systems, including two membrane-bound oxygen reductases, a bd-quinol oxidase and a cc(b/oo3 cytochrome oxidase. Effect of constant low oxygen concentration on growth and morphology of the wild-type, single (Δbd, Δcox and double deletion (Δcoxbd mutant strains of the genes encoding these oxygen reductases was studied. When both wild-type and deletion mutant strains were cultured in lactate/sulfate medium under constant 0.02% O2 sparging, they were able to grow but the final biomasses and the growth yield were lower than that obtained under anaerobic conditions. At the end of the growth, lactate was not completely consumed and when conditions were then switched to anaerobic, growth resumed. Time-lapse microscopy revealed that a large majority of the cells were then able to divide (over 97% but the time to recover a complete division event was longer for single deletion mutant Δbd than for the three other strains. Determination of the molar growth yields on lactate suggested that a part of the energy gained from lactate oxidation was derived toward cells protection/repairing against oxidative conditions rather than biosynthesis, and that this part was higher in the single deletion mutant Δbd and, to a lesser extent, Δcox strains. Our data show that when DvH encounters oxidative conditions, it is able to stop growing and to rapidly resume growing when conditions are switched to anaerobic, suggesting that it enters active dormancy sate under oxidative conditions. We propose that the pyruvate-ferredoxin oxidoreductase (PFOR plays a central role in this phenomenon by reversibly switching from an oxidative-sensitive fully active state to an oxidative-insensitive inactive state. The oxygen reductases, and especially the bd-quinol oxidase, would have a crucial function by maintaining

  17. Spectropolarimetric Survey of Hydrogen-rich White Dwarf Stars

    Czech Academy of Sciences Publication Activity Database

    Kawka, Adela; Vennes, S.; Schmidt, G. D.; Wickramasinghe, D.T.; Koch, R.

    2007-01-01

    Roč. 654, č. 1 (2007), s. 499-520 ISSN 0004-637X R&D Projects: GA ČR GP205/05/P186 Institutional research plan: CEZ:AV0Z10030501 Keywords : magnetic fields * white dwarf s Subject RIV: BN - Astronomy, Celestial Mechanics, Astrophysics Impact factor: 6.405, year: 2007

  18. Superconductivity in hydrogen-rich materials at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Drozdov, Alexander

    2016-07-01

    A room temperature superconductor is probably one of the most desired systems in solid state physics. The highest critical temperature (T{sub c}) that has been achieved so far is in the copper oxide system: 133 kelvin (K) at ambient pressure ([82]Schilling et al. 1993) and 160 K under pressure ([42]Gao et al. 1994). The nature of superconductivity in the cuprates and in the recently discovered iron-based superconductor family (T{sub c}=57 K) is still not fully understood. In contrast, there is a class of superconductors which is well-described by the Bardeen, Cooper, Schrieffer (BCS) theory - conventional superconductors. Great efforts were spent in searching for high-temperature (T{sub c} > 77 K) conventional superconductor but only T{sub c} = 39 K has been reached in MgB2 ([68]Nagamatsu et al. 2001). BCS theory puts no bounds for T{sub c} as follows from Eliashberg's formulation of BCS theory. T{sub c} can be high, if there is a favorable combination of high-frequency phonons, strong electron-phonon coupling, and a high density of states. It does not predict however in which materials all three parameters are large. At least it gives a clear indication that materials with light elements are favorable as light elements provide high frequencies in the phonon spectrum. The lightest element is hydrogen, and Ashcroft made a first prediction that metallic hydrogen will be a high-temperature superconductor ([6]Ashcroft 1968). As pressure of hydrogen metallization was too high (about 400-500 GPa) for experimental techniques then he proposed that compounds dominated by hydrogen (hydrides) also might be good high temperature superconductors ([6]Ashcroft 1968; [7]Ashcroft 2004). A lot of the followed calculations supported this idea. T{sub c} in the range of 50-235 kelvin was predicted for many hydrides. Unfortunately, only a moderate T{sub c} of 17 kelvin has been observed experimentally ([27]Eremets et al. 2008) so far. A goal of the present work is to find a hydrogen-dominant material with higher T{sub c}.

  19. Hydrogen rich gas production by thermocatalytic decomposition of kenaf biomass

    Energy Technology Data Exchange (ETDEWEB)

    Irmak, Sibel; Oeztuerk, ilker [Department of Chemistry, Cukurova University, Arts and Sciences Faculty, Adana 01330 (Turkey)

    2010-06-15

    Kenaf (Hibiscus cannabinus L.), a well known energy crop and an annual herbaceous plant grows very fast with low lodging susceptibility was used as representative lignocellulosic biomass in the present work. Thermocatalytic conversions were performed by aqueous phase reforming (APR) of kenaf hydrolysates and direct gasification of solid biomass of kenaf using 5% Pt on activated carbon as catalyst. Hydrolysates used in APR experiments were prepared by solubilization of kenaf biomass in subcritical water under CO{sub 2} gas pressure. APR of kenaf hydrolysate with low molecular weight polysaccharides in the presence of the reforming catalyst produced more gas compared to the hydrolysate that had high molecular weight polysaccharides. APR experiments of kenaf biomass hydrolysates and glucose, which was used as a simplest biomass model compound, in the presence of catalyst produced various amounts of gas mixtures that consisted of H{sub 2}, CO, CO{sub 2}, CH{sub 4} and C{sub 2}H{sub 6}. The ratios of H{sub 2} to other gases produced were 0.98, 1.50 and 1.35 for 150 C and 250 C subcritical water-treated kenaf hydrolysates and glucose, respectively. These ratios indicated that more the degraded organic content of kenaf hydrolysate the better selectivity for hydrogen production. Although APR of 250 C-kenaf hydrolysate resulted in similar gas content and composition as glucose, the gas volume produced was three times higher in glucose feed. The use of solid kenaf biomass as starting feedstock in APR experiments resulted in less gas production since the activity of catalyst was lowered by solid biomass particles. (author)

  20. Combinatorial and conventional studies on new highly selective methanation catalysts for the removal of small amounts of CO from hydrogen-rich gas mixtures; Kombinatorische und konventionelle Untersuchungen zu neuen hochselektiven Methanisierungskatalysatoren zur Entfernung geringer Mengen an CO aus wasserstoffreichen Gasgemischen

    Energy Technology Data Exchange (ETDEWEB)

    Kraemer, Michael

    2008-04-15

    New tailor-made catalysts for the purification of hydrogen-rich reformates by the selective methanation of CO were developed using combinatorial methods. The optimization of the catalysts was achieved within 3 or 4 generations while Ni-based oxides generally proved most promising. Conventional validations confirmed the successive enhancement of the materials. All in all, a number of catalysts providing higher CO hydrogenation activities combined with a lower reactivity towards the undesired methanation of the excess CO{sub 2} in comparison to an industrial reference was discovered. The application of numerous characterization techniques to the optimization sequence Ni{sub 100} - Zr{sub 10}Ni{sub 90} - Re{sub 2}Zr{sub 10}Ni{sub 88} - Re{sub 0,6}Zr{sub 15}Ni{sub 84,4} resulted in the following model: The catalysts are present in their as-prepared state as mixed metal oxide, which is (partly) demixed during the reductive pretreatment. The resulting (Re)Ni-particles seem to represent the actual active component while ZrO{sub 2} could stabilize the dispersion. Alloying with Ni, Re seems to modify the surface of the catalyst in such a way that it only marginally interacts with CO{sub 2}. Solo methanation tests unambiguously reveal that the increase in selectivity is not connected to a competition between CO and CO{sub 2} for adsorption sites but is based on a loss of the intrinsic reactivity of the respective samples towards the methanation of CO{sub 2}. (orig.)

  1. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  2. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

  3. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  4. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  5. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  6. BWR fuel assembly bottom nozzle with one-way coolant flow valve

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

  7. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  8. APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2015-01-01

    Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

  9. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  10. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

  11. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  12. The chemistry of the X-7 (organic) loop coolant part I, May 1960 to April 1965

    International Nuclear Information System (INIS)

    Smee, J.L.

    1966-01-01

    The report describes in detail the X-7 coolant chemistry from the start of loop operation in May 1960 to April 1965. During this period the coolant was Santowax OM containing a nominal 30% high boilers or high molecular weight decomposition products. During the first few months of operation it became apparent that there wa.s a serious problem in the fouling of fuel element heat transfer surfaces. This was overcome by continuous purification of the coolant by Attapulgus clay and filters. Since clay purification has been in use, the fouling rate has been less than 0.2 μg.cm -2 .h -1 (10 μm per year), the target value for successful operation of an organic cooled power reactor. Control of the fouling promoter chlorine has been accomplished by completely excluding it from the vicinity of the loop. Any which does get into the coolant is removed by a bed of Mg ribbon and Pd pellets. Since such a bed has been in use, the Cl content of the coolant has been less than 3 ppm. Also given in this report are: (a) a brief history of the loop since its inception in 1959. (b) the effect of the clay column on the coolant chemistry. (c) a complete description of the current purification, degas and make-up circuits, (d) a summary of the coolant chemistry during all fuel irradiations. (author)

  13. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  14. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  15. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  16. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  17. A study of the loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Y.W.; Chung, M.K.; Kim, S.H.; Park, J.S.; Lee, C.B.; Kim, S.B.; Won, S.Y.; Cho, Y.R.

    1983-01-01

    The primary objectives of this project are: (1) To review the published information on LOCA/ECCS study (2) To investigate reflood phenomena and to provide necessary information for analytical model development (3) To modyfy and develop a reflood analysis code. To review the published information on LOCA/ECCS, heat transfer phenomena are divided into 4 regions. Heat transfer correlations published in the references are reviewed and classified according to the regions. To investigate reflood phenomena and to provide better modeling of reflood phenomena, experments have been carried out with an electrically heated 3x3 rod bundle. Heat flux and heat transfer coefficients at the hot surface have been determined from the experimental data by HTC program. The influences of the parameters such as flooding rate, coolant subcooling and power generation on the propagation of rewetting front were also investigated. Calculations obtained from REFLUX code were compared with the experimental data to help an understanding of the reflood heat transfer mechanisms, and then some modifications of the code were provided. Improvements in heat transfer correlations of transition and inverted annular film boiling region, and the logic for the selection of heat transfer regime allowed better estimate for rod temperature behavior. (Author)

  18. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  19. ESBWR long term containment response to loss of coolant accidents

    International Nuclear Information System (INIS)

    Alamgir, M. D.; Marquino, W.; Diaz-Quiroz, J.; Tucker, L.

    2010-01-01

    ESBWR is a 4500 MWt generation III+ natural circulation reactor with an array of robust passive safety systems to keep the reactor safe during postulated transients and accidents. With the submittal of the latest revision of the Design Control Document (DCD) to US Nuclear Regulatory Commission, ESBWR is nearing the completion of the US certification process. This paper focuses on the bounding licensing analysis of the long-term (30-day) response of the ESBWR containment to limiting Loss of Coolant Accident (LOCA) performed with the TRACG code. It is shown that using only passive systems available during the first 72 hours after the limiting Main Steam Line Break LOCA, the predicted peak containment pressure in the ESBWR containment remain well below the design limits with good margin. After 72 hours of LOCA initiation, PCCS Vent Fans (non-safety system) become available that remove non-condensable gases from, and further enhance the effectiveness of, PCCS heat exchangers to reduce the containment pressure and temperature to values substantially below the design limits. During the post- 72 hour period, the beneficial effects of the Vent Fan operation, combined with the available operator action to refill of PCCS pools, continue to maintain the containment pressure to about 30% below the design limit at 30 days after a limiting ESBWR LOCA. (authors)

  20. Filtering device for primary coolant circuits in BWR type reactors

    International Nuclear Information System (INIS)

    Tajima, Fumio; Yamamoto, Tetsuo.

    1985-01-01

    Purpose: To obtain a filtering device with a large filtering area and requiring less space. Constitution: A condensate inlet for introducing condensates to be filtered of primary coolant circuits, a filtrate exit, a backwash water exit and a bent tube are disposed to a container, and a plurality of hollow thread membrane modules are suspended in the container. The condensates are caused to flow through the condensate inlet, filtered through the hollow thread membrane and then discharged from the filtrate exit. When the filtering treatment is proceeded to some extent, since solid contents captured in the hollow thread membranes are accumulated, a differential pressure is produced between the condensate inlet and the filtrate exit. When the differential pressure reaches a predetermined value, the backwash is conducted to discharge the liquid cleaning wastes through the backwash exit. The bent tube disposed to the container body is used for water and air draining. The hollow thread membranes are formed with porous resin such as of polyethylene. (Kawakami, Y.)

  1. Inspection of the Sizewll 'B' reactor coolant pump flywheels

    International Nuclear Information System (INIS)

    McNulty, A.L.; Cheshire, A.

    1992-01-01

    The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

  2. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  3. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  4. FILM-30: A Heat Transfer Properties Code for Water Coolant

    International Nuclear Information System (INIS)

    MARSHALL, THERON D.

    2001-01-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating

  5. Refurbishment of the IEAR1 primary coolant system piping supports

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

    2015-01-01

    A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

  6. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    International Nuclear Information System (INIS)

    Heising, Carolyn D.

    1998-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

  7. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    International Nuclear Information System (INIS)

    Patel, Bimal; Heising, C.D.

    1997-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specification limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (Author)

  8. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2002-01-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O 2 , H 2 , H 2 O 2 ), external solution composition (concentrations of O 2 , H 2 O 2 , and H 2 ), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  9. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y D; Choi, J H; Kim, T J; Jeong, K C; Kwon, S W; Kim, B H; Jeong, J Y; Park, J H; Kim, K R; Jo, B R

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  10. A study on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H 2 gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs

  11. Ageing of coolant channels in nuclear reactors (PHWRs)

    International Nuclear Information System (INIS)

    Mitra, T.L.; Chowdhury, M.K.; Gupta, R.K.; Pandarinathan, P.R.; Seth, V.K.

    1994-01-01

    In PHWRs, ageing of various components takes place due to factors like fast neutron flux, temperature, stress, environment etc. In coolant channel, the most severely affected component due to ageing is pressure tube, though other components like end fitting, calandria tube, garter spring spacer also show ageing to a limited extent. Ageing effects in pressure tube are seen in the form of diametral and axial creep, corrosion, delayed hydrogen cracking and irradiation hardening. In calandria tube and garter spring spacer, creep and hardening are seen though these are not of concern in PHWRs. In end fitting, irradiation embrittlement and abrasion of sealing faces are the areas of concern. Ageing process in these components are the areas of concern. Ageing process in these components are effectively retarded by taking measures like selection of proper material, manufacturing process, control of environmental chemistry, and design modifications. Experience and information gained in various Indian and foreign reactors have been used to improve upon the design in 220 MWe reactors and have formed the basis of design for 500 MWe reactors. (author). 3 refs., 5 figs

  12. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  13. On a specific feature of heat transfer to organic coolants

    International Nuclear Information System (INIS)

    Kafengauz, N.L.; Gladkikh, V.A.

    1986-01-01

    Heat transfer to organic coolants, which is accompanied by solid carbon deposit formation, is experimentally studied. Polished and rough steel tubes with 3 mm outside diameter and 0.5 mm wall thickness, heated by electric current, were used as fuel elements. Results of experiments with kerosene T-1 are presented under the following regime parameters: pressure - 45 b; flow rate - 3.75 m/s; temperature - 25-40 deg C; fuel element temperature - 400-900 deg C. In experiments on fuel elements with natural roughness deposit formation caused a smooth increase of the wall temperature. In fuel elements with polished surface, deposit formation caused during the first minutes the reduction of the wall temperature and after that it increased. Intensity of solid deposit formation in fuel elements with polished and rough surface was the same. Similar results were observed not only in experiments with kerosene T-1, but with other organic fluids as well: with toluene, n-heptane, diisopropylcyclohexane etc. The results obtained can be explained in the following way. Solid deposits on a smooth surface create roughness which improves heat exchange and reduces, respectively, the heating surface temperature. But deposits possess weak heat conductivity and create additional thermal resistance, which aggravates heat exchange. Interaction of these two factors causes the complicated time dependence of wall temperature

  14. Tables of thermodynamic properties of helium magnet coolant

    International Nuclear Information System (INIS)

    McAshan, M.

    1992-07-01

    The most complete treatment of the thermodynamic properties of helium at the present time is the monograph by McCarty: ''Thermodynamic Properties of Helium 4 from 2 to 1500 K at Pressures to 10 8 Pa'', Robert D. McCarty, Journal of Physical and Chemical Reference Data, Vol. 2, page 923--1040 (1973). In this work the complete range of data on helium is examined and the P-V-T surface is described by an equation of state consisting of three functions P(r,T) covering different regions together with rules for making the transition from one region to another. From this thermodynamic compilation together with correlations of the transport properties of helium was published the well-known NBS Technical Note: ''Thermophysical Properties of Helium 4 from 2 to 1500 K with pressures to 1000 Atmospheres'', Robert D. McCarty, US Department of Commerce, National Bureau of Standards Technical Note 631 (1972). This is the standard reference for helium cryogenics. The NBS 631 tables cover a wide range of temperature and pressure, and as a consequence, the number of points tabulated in the region of the single phase coolant for the SSC magnets are relatively few. The present work sets out to cover the range of interest in more detail in a way that is consistent with NBS 631. This new table is essentially identical to the older one and can be used as an auxiliary to it

  15. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

  16. Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR

    International Nuclear Information System (INIS)

    Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

    1983-01-01

    In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone

  17. A review of hydrodynamic instabilities and their relevance to mixing in molten fuel coolant interactions

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-03-01

    A review of the literature on Rayleigh-Taylor, Kelvin-Helmholtz and capillary instability is presented. The concept of Weber breakup is examined and found to involve a combination of the above instabilities. Sample calculations are given which show how these instabilities may contribute to the mixing of melt and coolant in a molten fuel coolant interaction. It is concluded that Rayleigh-Taylor instability is likely to be important as the melt falls into the coolant and that Kelvin-Helmholtz instability is likely to develop when significant vapour velocities occur. (author)

  18. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  19. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

  20. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  1. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  2. Recent results from the MIT in-core experiments on coolant chemistry

    International Nuclear Information System (INIS)

    Harling, O.K.; Kohse, G.E.; Cabello, E.C.; Bernard, J.A.

    1993-01-01

    This paper reports results from an ongoing series of in-core experiments that have been conducted at the 5-MW(thermal) MIT Research Reactor (MITR-II) for optimizing coolant chemistries in light water reactors. Four experiments are in progress, including a pressurized coolant chemistry loop (PCCL), a boiling coolant chemistry loop (BCCL), a facility for the study of irradiation-assisted stress-corrosion cracking, and one for the evaluation of in situ sensors for the monitoring of crack propagation in metal (SENSOR). The first two have now been fully operational for several years. The latter two are scheduled to begin regular operation later this year

  3. Qualitative infrared spectral analysis of products adsorbed by silica gel from ditolylmethane coolant and their adsorption isotherm

    International Nuclear Information System (INIS)

    Ermakov, V.A.; Benderskaya, O.S.

    1987-01-01

    The IR-spectral analysis has been applied to study the products adsorbed from ditolylmethane first-circuit coolant, as well as from still bottoms after coolant distillation on silicagel of various makes. The qualitative study of desorbate IR-spectra has shown that they refer to the classes of arylaldehydes, diarylketones and carbonic acids. Under actual conditions first-circuit reactor coolant also has a wide set of products of its radiolysis, therefore the spectrum of coolant oxidaton products must be wider. It is noted that adsorption on silica gel, ASK of oxygen-bearing compounds which are present in ditolyl methane coolant has 2 stages

  4. Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

    International Nuclear Information System (INIS)

    Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

    2007-01-01

    In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

  5. Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

    International Nuclear Information System (INIS)

    Melnikov, V.; Nigmatulin, B.

    1997-01-01

    The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

  6. Simulation of small break loss of coolant accident using relap 5/ MOD 2 computer code

    International Nuclear Information System (INIS)

    Megahed, M.M.

    1992-01-01

    An assessment of relap 5 / MOD 2/Cycle 36.05 best estimate computer code capabilities in predicting the thermohydraulic response of a PWR following a small break loss of coolant accident is presented. The experimental data base for the evaluation is the results of Test S-N H-3 performed in the semi scale MOD-2 c Test facility which modeled a 0.5% small break loss of coolant accident with an accompanying failure of the high pressure injection emergency core cooling system. A conclusion was reached that the code is capable of making small break loss of coolant accident calculations efficiently. However, some of the small break loss of coolant accident related phenomena were not properly predicted by the code, suggesting a need for code improvement.9 fig., 3 tab

  7. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  8. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  9. Radionuclide deposits on heat transfer surfaces in a circumt with dissociating N2O4 coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.; Komissarov, F.D.

    1984-01-01

    Radionuclides deposits on heat transfer surfaces of a circuit with dissociating coolant are studied. The areas of preferential deposition of 54 Mn, 51 Cr, 134 Cs and their distribution along the heating and cooling surfaces are determined. The comparison of the obtained data on the nuclide and chemical compositions of the deposits in the areas of N 2 O 4 coolant heating and cooling shows that 54 Mn, 51 Cr, 134 Cs deposit preferentially on heat transfer surfaces in the area of the coolant heating. Fixed and movable deposits consists of the structural material oxides. The quantity of radionuclides in the deposits on the surfaces of heat transfer tubes in the area of cooling decreases with the coolant temperature drop

  10. Prediction of thermal hydraulic parameters in the loss of coolant accident by using artificial neural networks

    International Nuclear Information System (INIS)

    Vaziri, N.; Erfani, A.; Monsefi, M.; Hajabri, A.

    2008-01-01

    In a reactor accident like loss of coolant accident , one or more signals may not be monitored by control panel for some reasons such as interruptions and so on. Therefore a fast alternative method could guarantee the safe and reliable exploration of nuclear power planets. In this study, we used artificial neural network with Elman recurrent structure to predict six thermal hydraulic signals in a loss of coolant accident after upper plenum break. In the prediction procedure, a few previous samples are fed to the artificial neural network and the output value or next time step is estimated by the network output. The Elman recurrent network is trained with the data obtained from the benchmark simulation of loss of coolant accident in VVER. The results reveal that the predicted values follow the real trends well and artificial neural network can be used as a fast alternative prediction tool in loss of coolant accident

  11. Confinement barriers for loss of coolant accidents in the SEAFP reactor plant models

    International Nuclear Information System (INIS)

    Blomquist, R.; Ebert, E.; Gay, J.M.; Mazille, F.; Natalizio, A.; Rolandsson, S.; Ross, W.E.; Shen, K.; Sjoeberg, A.

    1995-01-01

    Loss of coolant accidents may mobilise radioactivity and pressurise confinement barriers thereby making a release to the environment possible. The paper defines the radioactivity confinements and presents principal results from the underlying thermal-hydraulic analyses. (orig.)

  12. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  13. Effect of spacer grid mixing vanes on coolant outlet temperature distribution

    Energy Technology Data Exchange (ETDEWEB)

    Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

    2012-08-15

    In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

  14. Theoretical studying the stability of steady-state regime of a channel with a coolant condensation

    International Nuclear Information System (INIS)

    Savikhin, O.G.

    1987-01-01

    Based on the boiling channel stability theory, the channel steady-state stability with the coolant condensation is studied. Condensable coolants are used in the NPP steam-separator superheaters as well as in cryogenic technique. Under certain conditions the coolant flow rate and temperature fluctuations may be excited in the parallel channel system with coolant condensation, which produce a sufficient effect on the heat exchange equipment operation reliability. To describe unsteady processes of heat and mass transfer in the channel, a homogeneous two-phase flow one dimensional model is used. The results obtained allow one to make a conclusion concerning the effect of some parameters on condensing channel steady-state regime stability: reduction of inlet and outlet unheated communication length, pressure drop increase at the outlet plate and its reduction at the inlet one lead to the increase of stability margin

  15. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel

  16. Simulation of coolant mixing in pressure vessel reactors

    International Nuclear Information System (INIS)

    Hoehne, T.

    2003-06-01

    The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

  17. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  18. Reactor coolant pump testing using motor current signatures analysis

    Energy Technology Data Exchange (ETDEWEB)

    Burstein, N.; Bellamy, J.

    1996-12-01

    This paper describes reactor coolant pump motor testing carried out at Florida Power Corporation`s Crystal River plant using Framatome Technologies` new EMPATH (Electric Motor Performance Analysis and Trending Hardware) system. EMPATH{trademark} uses an improved form of Motor Current Signature Analysis (MCSA), technology, originally developed at Oak Ridge National Laboratories, for detecting deterioration in the rotors of AC induction motors. Motor Current Signature Analysis (MCSA) is a monitoring tool for motor driven equipment that provides a non-intrusive means for detecting the presence of mechanical and electrical abnormalities in the motor and the driven equipment. The base technology was developed at the Oak Ridge National Laboratory as a means for determining the affects of aging and service wear specifically on motor-operated valves used in nuclear power plant safety systems, but it is applicable to a broad range of electric machinery. MCSA is based on the recognition that an electric motor (ac or dc) driving a mechanical load acts as an efficient and permanently available transducer by sensing mechanical load variations, large and small, long-term and rapid, and converting them into variations in the induced current generated in the motor windings. The motor current variations, resulting from changes in load caused by gears, pulleys, friction, bearings, and other conditions that may change over the life of the motor, are carried by the electrical cables powering the motor and are extracted at any convenient location along the motor lead. These variations modulate the 60 Hz carrier frequency and appear as sidebands in the spectral plot.

  19. The electrochemistry of IGSCC mitigation in BWR coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

  20. Cladding embrittlement during postulated loss-of-coolant accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.; Yan, Y.; Burtseva, T.; Daum, R.; Nuclear Engineering Division

    2008-07-31

    The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at {le} 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

  1. Simulation of a loss of coolant accident with hydroaccumulator injection

    International Nuclear Information System (INIS)

    1988-10-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and organized two standard problem exercises. In these exercises, experimental data from the simulation of a 7.4% break loss of coolant accident was compared with analytical prediction of the behaviour of the facility calculated with computer codes. The second standard problem exercise involved a similar test, with the exception that in this case hydroaccumulator of the safety injection system were allowed to inject water in the system as anticipated in the design of the plant. This document presents a complete overview of the Second Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 22 refs, figs and tabs

  2. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Chen, N.C.J.; Yoder, G.L.; Wendel, M.W.

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs

  3. Fast measurements of the in-core coolant velocity in a BWR by neutron noise analysis

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der; Hoogenboom, J.E.

    1988-01-01

    A method to determine in-core coolant velocities from neutron noise within short time intervals has been developed. The accuracy of the method was determined by using a simulation set-up and by using signals of a twin self-powered neutron detector installed in the core of the Dodewaard BWR in the Netherlands. In-core coolant velocities can be estimated within 2.5 s with a standard deviation (due to statistics) less than 2.1%. The method is suitable for velocity monitoring as is shown by the application to a stepwise velocity change of the coolant in a model of a coolant channel of a BWR. The presented technique was applied to determine the variations of the coolant velocity in the Dodewaard core during normal operation and during pressure steps. Only minor variations of the coolant velocity were detected during normal reactor conditions. An increase of those variations with pressure lowering - indicating a lower thermal hydraulic stability - could be detected. A clear velocity response to pressure steps could be determined which was also reflected in the cross-spectrum of the velocity with the vessel pressure and with the in-core neutron flux. (author)

  4. Real-time reactor coolant system pressure/temperature limit system

    International Nuclear Information System (INIS)

    Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

    1991-01-01

    This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

  5. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  6. Method of suppressing the deposition of Co-60 to primary coolant pipeways in a nuclear reactor

    International Nuclear Information System (INIS)

    Hoshi, Michio; Tachikawa, Enzo; Goto, Satoshi; Sagawa, Chiaki; Yonezawa, Chushiro.

    1987-01-01

    Purpose: To suppress the deposition of Co-60 to primary coolant pipeways in a nuclear reactor. Method: To reduce the accumulation of Co-60 by causing chemical species of extremely similar chemical property with soluble Co-60 to be present together in coolants and replacing the deposition of Co-60 to the primary coolant pipeways in a nuclear reactor with that of the coexistent chemical spacies. Ni or Zn is used as the coexistet chemical spacies of similar chemical property with Co-60. The coexistent amount is from 5 to 10 times of the soluble Co-60 in the primary coolants. Ni or Zn solution adjusted with concentration is poured into and mixed with the coolants from a water feed source by using a high pressure constant volume pump. The amount of Co-60 taken into the pipeways caused by corrosion due to high temperature coolant is reduced to about 1/5 as compared with the case of Co-60 alone if 1 ppb of soluble Co-60 is present in water and 5 ppb of soluble Ni or Zn is added and, reduced to 1/12 if the amount of Ni or Zn is 10 ppb. (Kamimura, M.)

  7. Analysis of a water-coolant leak into a very high-temperature vitrification chamber

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

  8. Coolant material effect on the heat transfer rates of the molten metal pool with solidification

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1998-01-01

    Experimental studies on heat transfer and solidification of the molten metal pool with overlying coolant with boiling were performed. The simulant molten pool material is tin (Sn) with the melting temperature of 232 degree C. Demineralized water and R113 are used as the working coolant. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The Nusselt number and the Rayleigh number in the molten metal pool region of this study are compared between the water coolant case and the R113 coolant case. The experimental results for the water coolant are higher than those for R113. Also, the empirical relationship of the Nusselt number and the Rayleigh number is compared with the literature correlations measured from mercury. The present experimental results are higher than the literature correlations. It is believed that this discrepancy is caused by the effect of the heat loss to the environment on the natural convection heat transfer in the molten pool

  9. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  10. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    International Nuclear Information System (INIS)

    Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

    2017-01-01

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  11. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  12. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  13. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  14. Dryout heat flux in a debris bed with forced coolant flow from below

    International Nuclear Information System (INIS)

    Bang, Kwang-Hyun; Kim, Jong-Myung

    2004-01-01

    The objective of the present study is to experimentally investigate the enhancement of dryout heat flux in debris beds with coolant flow from below. The experimental facility consists mainly of an induction heater (40 kW, 35 kHz), a double-wall quartz-tube test section containing steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of particle bed was achieved by induction heating. This paper reports the experimental data for 5 mm particle bed and 300 mm bed height. The dryout heat rate data were obtained of both top-flooding case and forced coolant injection from below with the injection mass flux up to 1.5 kg/m 2 s. For the top-flooded case, the volumetric dryout heat rate was about 4 MW/m 3 and it increased as the rate of coolant injection from below was increased. At the coolant injection mass flux of 1.5 kg/m 2 s, the volumetric dryout heat rate was about 10 MW/m 3 , the enhancement factor was more than two. (author)

  15. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  16. Transient Temperature Distribution in a Reactor Core with Cylindrical Fuel Rods and Compressible Coolant

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-04-15

    Applying linearization and Laplace transformation the transient temperature distribution and weighted temperatures in fuel, canning and coolant are calculated analytically in two-dimensional cylindrical geometry for constant material properties in fuel and canning. The model to be presented includes previous models as special cases and has the following novel features: compressibility of the coolant is accounted for. The material properties of the coolant are variable. All quantities determining the temperature field are taken into account. It is shown that the solution for fuel and canning temperature may be given by the aid of 4 basic transfer functions depending on only two variables. These functions are calculated for all relevant rod geometries and material constants. The integrals involved in transfer functions determining coolant temperatures are solved for the most part generally by application of coordinate and Laplace transformation. The model was originally developed for use in steam cooled fast reactor analysis where the coolant temperature rise and compressibility are considerable. It may be applied to other fast or thermal systems after suitable simplifications.

  17. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  18. Design on Hygrometry System of Primary Coolant Circuit of HTR-PM

    International Nuclear Information System (INIS)

    Sun Yanfei; Zhong Shuoping; Huang Xiaojin

    2014-01-01

    Helium is the primary coolant in HTR-PM. If vapor get into the helium in primary coolant circuit because of some special reasons, such as the broken of steam-generator tube, chemical reaction will take effect between the graphite in reactor core and vapor in primary coolant circuit, and the safety of the reactor operation will be influenced. So the humidity of the helium in primary coolant circuit is one key parameter of HTR-PM to be monitored in-line. Once the humidity is too high, trigger signal of turning off the reactor must be issued. The hygrometry system of HTR-PM is consisting of filter, cooler, hygrometry sensor, flow meter, and some valves and tube. Helium with temperature of 250℃ is lead into the hygrometry system from the outlet of the main helium blower. After measuring, the helium is re-injected back to the primary circuit. No helium loses in this processing, and no other pump is needed. Key factors and calculations in design on hygrometry system of HTR-PM are described. A sample instrument has been made. Results of experiments proves that this hygrometry system is suitable for monitoring the humidity of the primary coolant of HTR-PM. (author)

  19. Coolant void effect investigation - case of a na-cooled fast reactor

    International Nuclear Information System (INIS)

    Glinatsis, G.; Gugiu, D.

    2013-01-01

    In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

  20. The solid coolant and prospects of its use in innovative reactors

    International Nuclear Information System (INIS)

    Dmitriev, A.M.; Deniskin, V.P.

    2010-01-01

    The progress of nuclear power demands consideration and development of innovative projects of the reactors having the increased level of safety due to their immanent properties allowing to provide high parameters. One of interesting and perspective offers is the use of a solid substance as a coolant. Use of the solid coolant of a nuclear reactor core has significant advantages among which an opportunity of movement of the coolant in the core under action of gravities and absence of necessity to have superfluous pressure in the jacket, that in turn means small metal consumption of construction, decrease in risk of emergency and its consequences. Cooling of the core with the help of solid substance is possible at performance of the certain conditions connected to features of the solid coolant. The major requirements are: the uniform continuous movement and minimal fluctuation of its density on every site of the core; high mechanical durability and wear resistance of particles; as well as good parameters of heat exchange, i.e. high heat conductivity and thermal capacity of the coolant material at the core operating conditions