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Sample records for akr-1 reactor

  1. AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

    International Nuclear Information System (INIS)

    Hansen, W.

    2003-01-01

    Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

  2. Enzymes of the AKR1B and AKR1C subfamilies and uterine diseases

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    Tea eLanisnik Rizner

    2012-03-01

    Full Text Available Endometrial and cervical cancers, uterine myoma, and endometriosis are very common uterine diseases. Worldwide, more than 800,000 women are affected annually by gynecological cancers, as a result of which, more than 360,000 die. During their reproductive age, about 70% of women develop uterine myomas, 10% to 15% suffer from endometriosis, and 35% to 50% from infertility associated with endometriosis. Uterine diseases are associated with aberrant inflammatory responses and concomitant increased production of prostaglandins (PG. They are also related to decreased differentiation, due to low levels of protective progesterone and retinoic acid, and to enhanced proliferation, due to high local concentrations of estrogens. The pathogenesis of these diseases can thus be attributed to disturbed PG, estrogen and retinoid metabolism and actions. Five human members of the aldo-keto reductase 1B (AKR1B and 1C (AKR1C superfamilies, i.e., AKR1B1, AKR1B10, AKR1C1, AKR1C2 and AKR1C3, have roles in these processes and can thus be implicated in uterine diseases. AKR1B1 and AKR1C3 catalyze the formation of PGF2alpha which stimulates cell proliferation. AKR1C3 converts PGD2 to 9alpha,11beta-PGF2, and thus counteracts the formation of 15deoxy-PGJ2, which can activate pro-apoptotic peroxisome-proliferator-activated receptor beta. AKR1B10 catalyzes the reduction of retinal to retinol, and in thus lessens the formation of retinoic acid, with potential pro-differentiating actions. The AKR1C1-AKR1C3 enzymes also act as 17-keto- and 20-ketosteroid reductases to varying extents, and are implicated in increased estradiol and decreased progesterone levels. This review comprises a short introduction to uterine diseases, followed by an overview of the current literature on the AKR1B and AKR1C expression in the uterus and in uterine diseases. The potential implications of the AKR1B and AKR1C enzymes and their pathophysiologies are then discussed, followed by conclusions and

  3. Pathophysiological roles of aldo-keto reductases (AKR1C1 and AKR1C3) in development of cisplatin resistance in human colon cancers.

    Science.gov (United States)

    Matsunaga, Toshiyuki; Hojo, Aki; Yamane, Yumi; Endo, Satoshi; El-Kabbani, Ossama; Hara, Akira

    2013-02-25

    Cisplatin (cis-diamminedichloroplatinum, CDDP) is widely used for treatment of patients with solid tumors formed in various organs including the lung, prostate and cervix, but is much less sensitive in colon and breast cancers. One major factor implicated in the ineffectiveness has been suggested to be acquisition of the CDDP resistance. Here, we established the CDDP-resistant phenotypes of human colon HCT15 cells by continuously exposing them to incremental concentrations of the drug, and monitored expressions of aldo-keto reductases (AKRs) 1A1, 1B1, 1B10, 1C1, 1C2 and 1C3. Among the six AKRs, AKR1C1 and AKR1C3 are highly induced with the CDDP resistance. The resistance lowered the sensitivity toward cellular damages evoked by oxidative stress-derived aldehydes, 4-hydroxy-2-nonenal and 4-oxo-2-nonenal that are detoxified by AKR1C1 and AKR1C3. Overexpression of AKR1C1 or AKR1C3 in the parental HCT15 cells mitigated the cytotoxicity of the aldehydes and CDDP. Knockdown of both AKR1C1 and AKR1C3 in the resistant cells or treatment of the cells with specific inhibitors of the AKRs increased the sensitivity to CDDP toxicity. Thus, the two AKRs participate in the mechanism underlying the CDDP resistance probably via detoxification of the aldehydes resulting from enhanced oxidative stress. The resistant cells also showed an enhancement in proteolytic activity of proteasome accompanied by overexpression of its catalytic subunits (PSMβ9 and PSMβ10). Pretreatment of the resistant cells with a potent proteasome inhibitor Z-Leu-Leu-Leu-al augmented the CDDP sensitization elicited by the AKR inhibitors. Additionally, the treatment of the cells with Z-Leu-Leu-Leu-al and the AKR inhibitors induced the expressions of the two AKRs and proteasome subunits. Collectively, these results suggest the involvement of up-regulated AKR1C1, AKR1C3 and proteasome in CDDP resistance of colon cancers and support a chemotherapeutic role for their inhibitors. Copyright © 2012 Elsevier Ireland

  4. Modeling single nucleotide polymorphisms in the human AKR1C1 and AKR1C2 genes: implications for functional and genotyping analyses.

    Directory of Open Access Journals (Sweden)

    Jonathan W Arthur

    2010-12-01

    Full Text Available Enzymes encoded by the AKR1C1 and AKR1C2 genes are responsible for the metabolism of progesterone and 5α-dihydrotestosterone (DHT, respectively. The effect of amino acid substitutions, resulting from single nucleotide polymorphisms (SNPs in the AKR1C2 gene, on the enzyme kinetics of the AKR1C2 gene product were determined experimentally by Takashi et al. In this paper, we used homology modeling to predict and analyze the structure of AKR1C1 and AKR1C2 genetic variants. The experimental reduction in enzyme activity in the AKR1C2 variants F46Y and L172Q, as determined by Takahashi et al., is predicted to be due to increased instability in cofactor binding, caused by disruptions to the hydrogen bonds between NADP and AKR1C2, resulting from the insertion of polar residues into largely non-polar environments near the site of cofactor binding. Other AKR1C2 variants were shown to involve either conservative substitutions or changes taking place on the surface of the molecule and distant from the active site, confirming the experimental finding of Takahashi et al. that these variants do not result in any statistically significant reduction in enzyme activity. The AKR1C1 R258C variant is predicted to have no effect on enzyme activity for similar reasons. Thus, we provide further insight into the molecular mechanism of the enzyme kinetics of these proteins. Our data also highlight previously reported difficulties with online databases.

  5. Pharmacogenetics of aldo-keto reductase 1C (AKR1C) enzymes.

    Science.gov (United States)

    Alshogran, Osama Y

    2017-10-01

    Genetic variation in metabolizing enzymes contributes to variable drug response and disease risk. Aldo-keto reductase type 1C (AKR1C) comprises a sub-family of reductase enzymes that play critical roles in the biotransformation of various drug substrates and endogenous compounds such as steroids. Several single nucleotide polymorphisms have been reported among AKR1C encoding genes, which may affect the functional expression of the enzymes. Areas covered: This review highlights and comprehensively discusses previous pharmacogenetic reports that have examined genetic variations in AKR1C and their association with disease development, drug disposition, and therapeutic outcomes. The article also provides information about the effect of AKR1C genetic variants on enzyme function in vitro. Expert opinion: The current evidence that links the effect of AKR1C gene polymorphisms to disease progression and development is inconsistent and needs further validation, despite of the tremendous knowledge available. Information about association of AKR1C genetic variants and drug efficacy, safety, and pharmacokinetics is limited, thus, future studies that advance our understanding about these relationships and their clinical relevance are needed. It is imperative to achieve consistent findings before the potential translation and adoption of AKR1C genetic variants in clinical practice.

  6. Hepatitis B Virus X Protein Up-Regulates AKR1C1 Expression Through Nuclear Factor-Y in Human Hepatocarcinoma Cells.

    Science.gov (United States)

    Li, Kai; Ding, Shijia; Chen, Ke; Qin, Dongdong; Qu, Jialin; Wang, Sen; Sheng, Yanrui; Zou, Chengcheng; Chen, Limin; Tang, Hua

    2013-01-01

    The hepatitis B virus X (HBx) protein has long been recognized as an important transcriptional transactivator of several genes. Human aldo-keto reductase family 1, member C1 (AKR1C1), a member of the family of AKR1CS, is significantly increased in HBx-expressed cells. This study aimed to investigate the possible mechanism of HBx in regulating AKR1C1 expression in HepG2.2.15 cells and the role of AKR1C1 for HBV-induced HCC. RT-PCR was performed to detect AKR1C1 expression on mRNA level in HepG2 and HepG2.2.15 cell. The promoter activity of AKR1C1 was assayed by transient transfection and Dual-luciferase reporter assay system. The AKR1C1 promoter sequence was screened using the TFSEARCH database and the ALIBABA 2.0 software. The potential transcription factors binding sites were identified using 5' functional deletion analysis and site-directed mutagenesis. In this study, we found that HBx promoted AKR1C1 expression in HepG2.2.15 cells. Knockdown of HBx inhibited AKR1C1 activation. The role of HBx expression in regulating the promoter activity of human AKR1C1 gene was analyzed. The 5'functional deletion analysis identified that the region between -128 and -88 was the minimal promoter region of HBx to activate AKR1C1 gene expression. Site-directed mutagenesis studies suggested that nuclear factor-Y (NF-Y) plays an important role in this HBx-induced AKR1C1 activation. In HepG2.2.1.5 cell, HBx can promote AKR1C1 promoter activity and thus activates the basal transcription of AKR1C1 gene. This process is mediated by the transcription factor NF-Y. This study explored the mechanism for the regulation of HBV on AKR1C1 expression and has provided a new understanding of HBV-induced HCC.

  7. High-resolution structure of AKR1a4 in the apo form and its interaction with ligands

    International Nuclear Information System (INIS)

    Faucher, Frédérick; Jia, Zongchao

    2012-01-01

    Despite its high affinity for NADPH, AKR1a4 crystallized in the apo form, which is very rare for aldo-keto reductase enzymes. Aldo-keto reductase 1a4 (AKR1a4; EC 1.1.1.2) is the mouse orthologue of human aldehyde reductase (AKR1a1), the founding member of the AKR family. As an NADPH-dependent enzyme, AKR1a4 catalyses the conversion of d-glucuronate to l-gulonate. AKR1a4 is involved in ascorbate biosynthesis in mice, but has also recently been found to interact with SMAR1, providing a novel mechanism of ROS regulation by ATM. Here, the crystal structure of AKR1a4 in its apo form at 1.64 Å resolution as well as the characterization of the binding of AKR1a4 to NADPH and P44, a peptide derived from SMAR1, is presented

  8. Selective reduction of AKR1C2 in prostate cancer and its role in DHT metabolism.

    Science.gov (United States)

    Ji, Qing; Chang, Lilly; VanDenBerg, David; Stanczyk, Frank Z; Stolz, Andrew

    2003-03-01

    As androgens play an essential role in prostate cancer, we sought to develop a real-time PCR to characterize mRNA expression profiles of human members of the Aldo-Keto Reductase (AKR) 1C gene family, as well as of 5 alpha-steroid reductase Type II (SRD5A2) in prostate cancer samples. Functional activity and regulation of AKR1C2, a 3 alpha-hydroxysteroid dehydrogenase (HSD) type III, was also assessed in prostate cancer cell lines. Gene specific PCR primers were established and relative gene expression of human AKR1C family members was determined in paired samples of cancerous and surrounding unaffected prostate tissue. AKR1C2 preferentially reduces DHT to the weak metabolite 5 alpha-androstane-3 alpha,17 beta-diol (3 alpha-diol) without conversion of 3 alpha-diol to DHT in the PC-3 cell line, and its expression was increased by DHT treatment in LNCaP cells. Selectively reduced expression of AKR1C2 mRNA, but not AKR1C1 (97% sequence identity), was found in approximately half of the pairs whereas AKR1C3 relative expression was not significantly altered. No aberrant expression of AKR1C4 expression or significant differences in SRD5A2 gene expression were found. AKR1C2 functions as a DHT reductase in prostate-derived cells lines and is regulated by DHT. Additional studies are needed to further define the significance of reduced AKR1C2 expression in prostate cancer and its potential role in modulating local availability of DHT. Copyright 2003 Wiley-Liss, Inc.

  9. AKR1B10 induces cell resistance to daunorubicin and idarubicin by reducing C13 ketonic group

    International Nuclear Information System (INIS)

    Zhong Linlin; Shen Honglin; Huang Chenfei; Jing, Hongwu; Cao Deliang

    2011-01-01

    Daunorubicin, idarubicin, doxorubicin and epirubicin are anthracyclines widely used for the treatment of lymphoma, leukemia, and breast, lung, and liver cancers, but tumor resistance limits their clinical success. Aldo-keto reductase family 1 B10 (AKR1B10) is an NADPH-dependent enzyme overexpressed in liver and lung carcinomas. This study was aimed to determine the role of AKR1B10 in tumor resistance to anthracyclines. AKR1B10 activity toward anthracyclines was measured using recombinant protein. Cell resistance to anthracycline was determined by ectopic expression of AKR1B10 or inhibition by epalrestat. Results showed that AKR1B10 reduces C13-ketonic group on side chain of daunorubicin and idarubicin to hydroxyl forms. In vitro, AKR1B10 converted daunorubicin to daunorubicinol at V max of 837.42 ± 81.39 nmol/mg/min, K m of 9.317 ± 2.25 mM and k cat /K m of 3.24. AKR1B10 showed better catalytic efficiency toward idarubicin with V max at 460.23 ± 28.12 nmol/mg/min, K m at 0.461 ± 0.09 mM and k cat /K m at 35.94. AKR1B10 was less active toward doxorubicin and epirubicin with a C14-hydroxyl group. In living cells, AKR1B10 efficiently catalyzed reduction of daunorubicin (50 nM) and idarubicin (30 nM) to corresponding alcohols. Within 24 h, approximately 20 ± 2.7% of daunorubicin (1 μM) or 23 ± 2.3% of idarubicin (1 μM) was converted to daunorubicinol or idarubicinol in AKR1B10 expression cells compared to 7 ± 0.9% and 5 ± 1.5% in vector control. AKR1B10 expression led to cell resistance to daunorubicin and idarubicin, but inhibitor epalrestat showed a synergistic role with these agents. Together our data suggest that AKR1B10 participates in cellular metabolism of daunorubicin and idarubicin, resulting in drug resistance. These data are informative for the clinical use of idarubicin and daunorubicin. - Highlights: → This study defines enzyme activity of AKR1B10 protein towards daunorubicin, idarubicin, doxorubicin, and epirubicin. → This study pinpoints

  10. Purification and characterization of akr1b10 from human liver: role in carbonyl reduction of xenobiotics.

    Science.gov (United States)

    Martin, Hans-Jörg; Breyer-Pfaff, Ursula; Wsol, Vladimir; Venz, Simone; Block, Simone; Maser, Edmund

    2006-03-01

    Members of the aldo-keto reductase (AKR) superfamily have a broad substrate specificity in catalyzing the reduction of carbonyl group-containing xenobiotics. In the present investigation, a member of the aldose reductase subfamily, AKR1B10, was purified from human liver cytosol. This is the first time AKR1B10 has been purified in its native form. AKR1B10 showed a molecular mass of 35 kDa upon gel filtration and SDS-polyacrylamide gel electrophoresis. Kinetic parameters for the NADPH-dependent reduction of the antiemetic 5-HT3 receptor antagonist dolasetron, the antitumor drugs daunorubicin and oracin, and the carcinogen 4-methylnitrosamino-1-(3-pyridyl)-1-butanone (NNK) to the corresponding alcohols have been determined by HPLC. Km values ranged between 0.06 mM for dolasetron and 1.1 mM for daunorubicin. Enzymatic efficiencies calculated as kcat/Km were more than 100 mM-1 min-1 for dolasetron and 1.3, 0.43, and 0.47 mM-1 min-1 for daunorubicin, oracin, and NNK, respectively. Thus, AKR1B10 is one of the most significant reductases in the activation of dolasetron. In addition to its reducing activity, AKR1B10 catalyzed the NADP+-dependent oxidation of the secondary alcohol (S)-1-indanol to 1-indanone with high enzymatic efficiency (kcat/Km=112 mM-1 min-1). The gene encoding AKR1B10 was cloned from a human liver cDNA library and the recombinant enzyme was purified. Kinetic studies revealed lower activity of the recombinant compared with the native form. Immunoblot studies indicated large interindividual variations in the expression of AKR1B10 in human liver. Since carbonyl reduction of xenobiotics often leads to their inactivation, AKR1B10 may play a role in the occurrence of chemoresistance of tumors toward carbonyl group-bearing cytostatic drugs.

  11. The aldo-keto reductase AKR1B7 coexpresses with renin without influencing renin production and secretion.

    Science.gov (United States)

    Machura, Katharina; Iankilevitch, Elina; Neubauer, Björn; Theuring, Franz; Kurtz, Armin

    2013-03-01

    On the basis of evidence that within the adult kidney, the aldo-keto reductase AKR1B7 (aldo-keto reductase family 1, member 7, also known as mouse vas deferens protein, MVDP) is selectively expressed in renin-producing cells, we aimed to define a possible role of AKR1B7 for the regulation and function of renin cells in the kidney. We could confirm colocalization and corecruitment of renin and of AKR1B7 in wild-type kidneys. Renin cells in AKR1B7-deficient kidneys showed normal morphology, numbers, and intrarenal distribution. Plasma renin concentration (PRC) and renin mRNA levels of AKR1B7-deficient mice were normal at standard chow and were lowered by a high-salt diet directly comparable to wild-type mice. Treatment with a low-salt diet in combination with an angiotensin-converting enzyme inhibitor strongly increased PRC and renin mRNA in a similar fashion both in AKR1B7-deficient and wild-type mice. Under this condition, we also observed a strong retrograde recruitment of renin-expressing cell along the preglomerular vessels, however, without a difference between AKR1B7-deficient and wild-type mice. The isolated perfused mouse kidney model was used to study the acute regulation of renin secretion by ANG II and by perfusion pressure. Regarding these parameters, no differences were observed between AKR1B7-deficient and wild-type kidneys. In summary, our data suggest that AKR1B7 is not of major relevance for the regulation of renin production and secretion in spite of its striking coregulation with renin expression.

  12. Overcoming CRPC Treatment Resistance via Novel Dual AKR1C3 Targeting

    Science.gov (United States)

    2017-10-01

    Accomplishments 3 4. Impact 8 5. Changes/Problems 8 6. Products 8 7. Participants & Other Collaborating Organizations 8 8. Special Reporting Requirements...orthotopic animal model 3. ACCOMPLISHMENTS: What were the major goals of the project? Specific Aim 1 To determine the mechanisms of AKR1C3-mediated...selected for stable clones with respective antibiotics . Here we showed the characterization of the tet-inducible LNCaP/TR/AKR1C3 stale clone

  13. Nuclear technology education at the new AKR-2 of the technical university Dresden

    International Nuclear Information System (INIS)

    Hansen, W.; Wolf, T.; Hurtado, A.

    2009-01-01

    The former research and training reactor AKR-1 was completely renewed, including the peripheral technical systems and the modernization of the reactor instrumentation with digital control technology. After licensing by the local authorities the technical University Dresden has Germany's latest training reactor. Basic experiments are performed for the following disciplines: nuclear energy technology, physics, teacher training, industrial engineering, nuclear medicine. Training courses cover nuclear medicine, nuclear physics, radiation protection and reactor physics. Further tasks include research program on neutron detectors, neutron physics, radiation spectroscopy, nuclear data bases.

  14. Reductive detoxification of acrolein as a potential role for aldehyde reductase (AKR1A) in mammals.

    Science.gov (United States)

    Kurahashi, Toshihiro; Kwon, Myoungsu; Homma, Takujiro; Saito, Yuka; Lee, Jaeyong; Takahashi, Motoko; Yamada, Ken-Ichi; Miyata, Satoshi; Fujii, Junichi

    2014-09-12

    Aldehyde reductase (AKR1A), a member of the aldo-keto reductase superfamily, suppresses diabetic complications via a reduction in metabolic intermediates; it also plays a role in ascorbic acid biosynthesis in mice. Because primates cannot synthesize ascorbic acid, a principle role of AKR1A appears to be the reductive detoxification of aldehydes. In this study, we isolated and immortalized mouse embryonic fibroblasts (MEFs) from wild-type (WT) and human Akr1a-transgenic (Tg) mice and used them to investigate the potential roles of AKR1A under culture conditions. Tg MEFs showed higher methylglyoxal- and acrolein-reducing activities than WT MEFs and also were more resistant to cytotoxicity. Enzymatic analyses of purified rat AKR1A showed that the efficiency of the acrolein reduction was about 20% that of glyceraldehyde. Ascorbic acid levels were quite low in the MEFs, and while the administration of ascorbic acid to the cells increased the intracellular levels of ascorbic acid, it had no affect on the resistance to acrolein. Endoplasmic reticulum stress and protein carbonylation induced by acrolein treatment were less evident in Tg MEFs than in WT MEFs. These data collectively indicate that one of the principle roles of AKR1A in primates is the reductive detoxification of aldehydes, notably acrolein, and protection from its detrimental effects. Copyright © 2014 Elsevier Inc. All rights reserved.

  15. Elevated AKR1C3 expression promotes prostate cancer cell survival and prostate cell-mediated endothelial cell tube formation: implications for prostate cancer progressioan

    International Nuclear Information System (INIS)

    Dozmorov, Mikhail G; Lin, Hsueh-Kung; Azzarello, Joseph T; Wren, Jonathan D; Fung, Kar-Ming; Yang, Qing; Davis, Jeffrey S; Hurst, Robert E; Culkin, Daniel J; Penning, Trevor M

    2010-01-01

    Aldo-keto reductase (AKR) 1C family member 3 (AKR1C3), one of four identified human AKR1C enzymes, catalyzes steroid, prostaglandin, and xenobiotic metabolism. In the prostate, AKR1C3 is up-regulated in localized and advanced prostate adenocarcinoma, and is associated with prostate cancer (PCa) aggressiveness. Here we propose a novel pathological function of AKR1C3 in tumor angiogenesis and its potential role in promoting PCa progression. To recapitulate elevated AKR1C3 expression in cancerous prostate, the human PCa PC-3 cell line was stably transfected with an AKR1C3 expression construct to establish PC3-AKR1C3 transfectants. Microarray and bioinformatics analysis were performed to identify AKR1C3-mediated pathways of activation and their potential biological consequences in PC-3 cells. Western blot analysis, reverse transcription-polymerase chain reaction (RT-PCR), enzyme-linked immunosorbent assay (ELISA), and an in vitro Matrigel angiogenesis assays were applied to validate the pro-angiogenic activity of PC3-AKR1C3 transfectants identified by bioinformatics analysis. Microarray and bioinformatics analysis suggested that overexpression of AKR1C3 in PC-3 cells modulates estrogen and androgen metabolism, activates insulin-like growth factor (IGF)-1 and Akt signaling pathways, as well as promotes tumor angiogenesis and aggressiveness. Levels of IGF-1 receptor (IGF-1R) and Akt activation as well as vascular endothelial growth factor (VEGF) expression and secretion were significantly elevated in PC3-AKR1C3 transfectants in comparison to PC3-mock transfectants. PC3-AKR1C3 transfectants also promoted endothelial cell (EC) tube formation on Matrigel as compared to the AKR1C3-negative parental PC-3 cells and PC3-mock transfectants. Pre-treatment of PC3-AKR1C3 transfectants with a selective IGF-1R kinase inhibitor (AG1024) or a non-selective phosphoinositide 3-kinases (PI3K) inhibitor (LY294002) abolished ability of the cells to promote EC tube formation. Bioinformatics

  16. Novel SNPs of WNK1 and AKR1C3 are associated with preeclampsia.

    Science.gov (United States)

    Sun, Cheng-Juan; Li, Lin; Li, Xueyan; Zhang, Wei-Yuan; Liu, Xiao-Wei

    2018-08-20

    Preeclampsia is a hypertensive disorder of pregnancy and is one of the most common causes of poor perinatal outcomes. Preeclampsia increases the risk of hypertension in the future. Variants of WNK1 (lysine deficient protein kinase 1), ADRB2 (β2 adrenergic receptor), NEDD4L (ubiquitin-protein ligase NEDD4-like), KLK1 (kallikrein 1) contribute to hypertension, and AKR1C3 (aldo-keto reductase family1 member C3), is associated with preeclampsia. The association of single nucleotide polymorphisms (SNPs) in these five candidate preeclampsia susceptibility genes and the related traits in Chinese individuals were investigated. In this study, 13 SNPs of the five genes were genotyped in 276 preeclampsia patients and 229 age- and area-matched normal pregnancies in women of Chinese Northern Han origin. The 95% confidence interval (CI) and odds ratio (OR) were estimated by binary logistic regression. No obvious linkage disequilibrium or haplotypes were observed among these SNPs. Those with GG genotype and allele G of AKR1C3 (rs10508293) had a decreased risk of preeclampsia (adjusted OR = 3.011, 95% CI = 1.758-5.159, and adjusted OR = 1.745, 95% CI = 1.349-2.257, respectively). The AA genotype and allele A of WNK1 (rs1468326) were significantly associated with an increased risk in preeclampsia (adjusted OR = 2.307, 95% CI = 1.206-3.443, and adjusted OR = 1.663, 95% CI = 1.283-2.157, respectively). The findings indicate that the GG genotype of AKR1C3 rs10508293 is associated with decreased risk for preeclampsia and the AA genotype of WNK1 rs1468326 are related with an increased risk for preeclampsia. Copyright © 2018 Elsevier B.V. All rights reserved.

  17. Effect of allogenic thymic cells on radioleukaemogenesis in AKR-T1ALD mice

    International Nuclear Information System (INIS)

    Legrand, E.; Sankar-Mistry, P.; Kressmann, M.C.

    1975-01-01

    When AKR mice are irradiated with a sub-lethal dose (4 times 175 R), thymic lymphosarcomas (L.S.) occur earlier than in controls. This accelerated leukaemogenesis is not inhibited by syngenic restoration with bone marrows cells (BM). Using the AKR/T1ALD substrain which bears 38 chromosomes with 1 metacentric markers, it has been shown that AKR radio-chimaeras restored by T1ALD BM developed two kinds of L.S.: early (radiation-induced) L.S. originating mainly from host cells surviving irradiation and late L.S. from donor cells. The experiments were to investigate the potential influence of normal allogenic thymic cells, with or without syngenic B.M., on the incidence, latency and origin of LS appearing in irradiated AKR recipients. Adding C3H allogenic thymic cells to syngenic B.M. increases the percentage of early L.S. whose latencies are unchanged. Besides, when C3H thymic cells are injected to irradiated controls without syngenic B.M. cells, L.S. are seen to occur significantly earlier than in just the irradiated animals alone. In radio-chimaeras restored by allogenic thymic cells and syngenic B.M., except in one case, all the L.S. were seen to originate from B.M. cells. The interpretation of these results depends on the possible role of allogenic thymic cells on host cells surviving the irradiation, or the exogeneous B.M. In the first case, allogenic thymocytes could induce a graft versus host reaction increasing the post-irradiation depletion of lymphoid system and hastening thymic endoregeneration which is supposed to be the first step towards leukaemogenesis. The second hypothesis, which seems the most likely, would be that C1H thymic cells could selectively act on host cells surviving irradiation and enhance the differenciation of haemopoietic precursors at the expense of the lymphoid cells [fr

  18. Depressed levels of prostaglandin F2α in mice lacking Akr1b7 increase basal adiposity and predispose to diet-induced obesity.

    Science.gov (United States)

    Volat, Fanny E; Pointud, Jean-Christophe; Pastel, Emilie; Morio, Béatrice; Sion, Benoit; Hamard, Ghislaine; Guichardant, Michel; Colas, Romain; Lefrançois-Martinez, Anne-Marie; Martinez, Antoine

    2012-11-01

    Negative regulators of white adipose tissue (WAT) expansion are poorly documented in vivo. Prostaglandin F(2α) (PGF(2α)) is a potent antiadipogenic factor in cultured preadipocytes, but evidence for its involvement in physiological context is lacking. We previously reported that Akr1b7, an aldo-keto reductase enriched in adipose stromal vascular fraction but absent from mature adipocytes, has antiadipogenic properties possibly supported by PGF(2α) synthase activity. To test whether lack of Akr1b7 could influence WAT homeostasis in vivo, we generated Akr1b7(-/-) mice in 129/Sv background. Akr1b7(-/-) mice displayed excessive basal adiposity resulting from adipocyte hyperplasia/hypertrophy and exhibited greater sensitivity to diet-induced obesity. Following adipose enlargement and irrespective of the diet, they developed liver steatosis and progressive insulin resistance. Akr1b7 loss was associated with decreased PGF(2α) WAT contents. Cloprostenol (PGF(2α) agonist) administration to Akr1b7(-/-) mice normalized WAT expansion by affecting both de novo adipocyte differentiation and size. Treatment of 3T3-L1 adipocytes and Akr1b7(-/-) mice with cloprostenol suggested that decreased adipocyte size resulted from inhibition of lipogenic gene expression. Hence, Akr1b7 is a major regulator of WAT development through at least two PGF(2α)-dependent mechanisms: inhibition of adipogenesis and lipogenesis. These findings provide molecular rationale to explore the status of aldo-keto reductases in dysregulations of adipose tissue homeostasis.

  19. Kinetic alteration of a human dihydrodiol/3alpha-hydroxysteroid dehydrogenase isoenzyme, AKR1C4, by replacement of histidine-216 with tyrosine or phenylalanine.

    Science.gov (United States)

    Ohta, T; Ishikura, S; Shintani, S; Usami, N; Hara, A

    2000-01-01

    Human dihydrodiol dehydrogenase with 3alpha-hydroxysteroid dehydrogenase activity exists in four forms (AKR1C1-1C4) that belong to the aldo-keto reductase (AKR) family. Recent crystallographic studies on the other proteins in this family have indicated a role for a tyrosine residue (corresponding to position 216 in these isoenzymes) in stacking the nicotinamide ring of the coenzyme. This tyrosine residue is conserved in most AKR family members including AKR1C1-1C3, but is replaced with histidine in AKR1C4 and phenylalanine in some AKR members. In the present study we prepared mutant enzymes of AKR1C4 in which His-216 was replaced with tyrosine or phenylalanine. The two mutations decreased 3-fold the K(m) for NADP(+) and differently influenced the K(m) and k(cat) for substrates depending on their structures. The kinetic constants for bile acids with a 12alpha-hydroxy group were decreased 1.5-7-fold and those for the other substrates were increased 1.3-9-fold. The mutation also yielded different changes in sensitivity to competitive inhibitors such as hexoestrol analogues, 17beta-oestradiol, phenolphthalein and flufenamic acid and 3,5,3', 5'-tetraiodothyropropionic acid analogues. Furthermore, the mutation decreased the stimulatory effects of the enzyme activity by sulphobromophthalein, clofibric acid and thyroxine, which increased the K(m) for the coenzyme and substrate of the mutant enzymes more highly than those of the wild-type enzyme. These results indicate the importance of this histidine residue in creating the cavity of the substrate-binding site of AKR1C4 through the orientation of the nicotinamide ring of the coenzyme, as well as its involvement in the conformational change by binding non-essential activators. PMID:11104674

  20. AKR1C3-Mediated Adipose Androgen Generation Drives Lipotoxicity in Women With Polycystic Ovary Syndrome.

    Science.gov (United States)

    O'Reilly, Michael W; Kempegowda, Punith; Walsh, Mark; Taylor, Angela E; Manolopoulos, Konstantinos N; Allwood, J William; Semple, Robert K; Hebenstreit, Daniel; Dunn, Warwick B; Tomlinson, Jeremy W; Arlt, Wiebke

    2017-09-01

    Polycystic ovary syndrome (PCOS) is a prevalent metabolic disorder occurring in up to 10% of women of reproductive age. PCOS is associated with insulin resistance and cardiovascular risk. Androgen excess is a defining feature of PCOS and has been suggested as causally associated with insulin resistance; however, mechanistic evidence linking both is lacking. We hypothesized that adipose tissue is an important site linking androgen activation and metabolic dysfunction in PCOS. We performed a human deep metabolic in vivo phenotyping study examining the systemic and intra-adipose effects of acute and chronic androgen exposure in 10 PCOS women, in comparison with 10 body mass index-matched healthy controls, complemented by in vitro experiments. PCOS women had increased intra-adipose concentrations of testosterone (P = 0.0006) and dihydrotestosterone (P = 0.01), with increased expression of the androgen-activating enzyme aldo-ketoreductase type 1 C3 (AKR1C3) (P = 0.04) in subcutaneous adipose tissue. Adipose glycerol levels in subcutaneous adipose tissue microdialysate supported in vivo suppression of lipolysis after acute androgen exposure in PCOS (P = 0.04). Mirroring this, nontargeted serum metabolomics revealed prolipogenic effects of androgens in PCOS women only. In vitro studies showed that insulin increased adipose AKR1C3 expression and activity, whereas androgen exposure increased adipocyte de novo lipid synthesis. Pharmacologic AKR1C3 inhibition in vitro decreased de novo lipogenesis. These findings define an intra-adipose mechanism of androgen activation that contributes to adipose remodeling and a systemic lipotoxic metabolome, with intra-adipose androgens driving lipid accumulation and insulin resistance in PCOS. AKR1C3 represents a promising therapeutic target in PCOS. Copyright © 2017 Endocrine Society

  1. Characteristics of AKR sources: A statistical description

    International Nuclear Information System (INIS)

    Hilgers, A.; Roux, A.; Lundin, R.

    1991-01-01

    A description of plasma properties within the sources of the Auroral Kilometric Radiation (AKR) is given. It is based on data collected during ∼ 50 AKR source crossings in the altitude range between 4,000 and 9,000 km by the Swedish spacecraft Viking. The following results are obtained; (i) the frequency of the lowest frequency peak of the AKR f peak is found to be very close to f ce , the electron gyrofrequency ((f peak -f ce )/f ce ≤ 0.08), on the average, (ii) the lower cutoff frequency f LC is on the average at f ce ((f LC -f ce )/f ce ≅ 0), (iii) in the sources the density is typically less than 1.5 cm -3 , which is of the order of the density of hot electrons and (iv) the source is located within an acceleration region, as evidenced by electrons accelerated above and ions accelerated below

  2. Novel chemical scaffolds of the tumor marker AKR1B10 inhibitors discovered by 3D QSAR pharmacophore modeling.

    Science.gov (United States)

    Kumar, Raj; Son, Minky; Bavi, Rohit; Lee, Yuno; Park, Chanin; Arulalapperumal, Venkatesh; Cao, Guang Ping; Kim, Hyong-ha; Suh, Jung-keun; Kim, Yong-seong; Kwon, Yong Jung; Lee, Keun Woo

    2015-08-01

    Recent evidence suggests that aldo-keto reductase family 1 B10 (AKR1B10) may be a potential diagnostic or prognostic marker of human tumors, and that AKR1B10 inhibitors offer a promising choice for treatment of many types of human cancers. The aim of this study was to identify novel chemical scaffolds of AKR1B10 inhibitors using in silico approaches. The 3D QSAR pharmacophore models were generated using HypoGen. A validated pharmacophore model was selected for virtual screening of 4 chemical databases. The best mapped compounds were assessed for their drug-like properties. The binding orientations of the resulting compounds were predicted by molecular docking. Density functional theory calculations were carried out using B3LYP. The stability of the protein-ligand complexes and the final binding modes of the hit compounds were analyzed using 10 ns molecular dynamics (MD) simulations. The best pharmacophore model (Hypo 1) showed the highest correlation coefficient (0.979), lowest total cost (102.89) and least RMSD value (0.59). Hypo 1 consisted of one hydrogen-bond acceptor, one hydrogen-bond donor, one ring aromatic and one hydrophobic feature. This model was validated by Fischer's randomization and 40 test set compounds. Virtual screening of chemical databases and the docking studies resulted in 30 representative compounds. Frontier orbital analysis confirmed that only 3 compounds had sufficiently low energy band gaps. MD simulations revealed the binding modes of the 3 hit compounds: all of them showed a large number of hydrogen bonds and hydrophobic interactions with the active site and specificity pocket residues of AKR1B10. Three compounds with new structural scaffolds have been identified, which have stronger binding affinities for AKR1B10 than known inhibitors.

  3. AKR1C1 as a Biomarker for Differentiating the Biological Effects of Combustible from Non-Combustible Tobacco Products.

    Science.gov (United States)

    Woo, Sangsoon; Gao, Hong; Henderson, David; Zacharias, Wolfgang; Liu, Gang; Tran, Quynh T; Prasad, G L

    2017-05-03

    Smoking has been established as a major risk factor for developing oral squamous cell carcinoma (OSCC), but less attention has been paid to the effects of smokeless tobacco products. Our objective is to identify potential biomarkers to distinguish the biological effects of combustible tobacco products from those of non-combustible ones using oral cell lines. Normal human gingival epithelial cells (HGEC), non-metastatic (101A) and metastatic (101B) OSCC cell lines were exposed to different tobacco product preparations (TPPs) including cigarette smoke total particulate matter (TPM), whole-smoke conditioned media (WS-CM), smokeless tobacco extract in complete artificial saliva (STE), or nicotine (NIC) alone. We performed microarray-based gene expression profiling and found 3456 probe sets from 101A, 1432 probe sets from 101B, and 2717 probe sets from HGEC to be differentially expressed. Gene Set Enrichment Analysis (GSEA) revealed xenobiotic metabolism and steroid biosynthesis were the top two pathways that were upregulated by combustible but not by non-combustible TPPs. Notably, aldo-keto reductase genes, AKR1C1 and AKR1C2 , were the core genes in the top enriched pathways and were statistically upregulated more than eight-fold by combustible TPPs. Quantitative real time polymerase chain reaction (qRT-PCR) results statistically support AKR1C1 as a potential biomarker for differentiating the biological effects of combustible from non-combustible tobacco products.

  4. Differential expression patterns of Nqo1, AKR1B8 and Ho-1 in the liver and small intestine of C57BL/6 mice treated with sulforaphane

    Directory of Open Access Journals (Sweden)

    Lin Luo

    2015-12-01

    Full Text Available This data article contains complementary figures and results related to the research article entitled “butylated hydroxyanisole induces distinct expression patterns of Nrf2 and detoxification enzymes in the liver and small intestine of C57BL/6 mice” (Luo et al., 2015 [1], which defined the basal and butylated hydroxyanisole (BHA-induced expression patterns of Phase II enzymes Nqo1, AKR1B8, and Ho-1 in the liver and small intestine of C57BL/6 mice. Sulforaphane [1-isothiocyanato-4-(methylsulfinylbutane] (SFN, a naturally occurring isothiocyanate derived from cruciferous vegetables, is a highly potent inducer of phase II cytoprotective enzymes. This dataset reports the histological changes of Nqo1, AKR1B8, and Ho-1 in wild-type (WT and Nrf2-/- mice induced by SFN. The mice were given a 25 mg/kg single oral dose of SFN for 24 h and 48 h. Immunohistochemistry revealed that, in the liver from WT mice, SFN increased Nqo1 staining in hepatocytes with slight higher staining in the pericentral region. The induction of AKR1B8 appeared mostly in hepatocytes in the periportal region. The basal and inducible Ho-1 was located predominately in Kupffer cells. In the small intestine from WT mice, the inducible expression of Nqo1 and AKR1B8 appeared more obvious in the villus than that in the crypt.

  5. Sulindac inhibits pancreatic carcinogenesis in LSL-KrasG12D-LSL-Trp53R172H-Pdx-1-Cre mice via suppressing aldo-keto reductase family 1B10 (AKR1B10).

    Science.gov (United States)

    Li, Haonan; Yang, Allison L; Chung, Yeon Tae; Zhang, Wanying; Liao, Jie; Yang, Guang-Yu

    2013-09-01

    Sulindac has been identified as a competitive inhibitor of aldo-keto reductase 1B10 (AKR1B10), an enzyme that plays a key role in carcinogenesis. AKR1B10 is overexpressed in pancreatic ductal adenocarcinoma (PDAC) and exhibits lipid substrate specificity, especially for farnesyl and geranylgeranyl. There have been no studies though showing that the inhibition of PDAC by sulindac is via inhibition of AKR1B10, particularly the metabolism of farnesyl/geranylgeranyl and Kras protein prenylation. To determine the chemopreventive effects of sulindac on pancreatic carcinogenesis, 5-week-old LSL-Kras(G12D)-LSL-Trp53(R172H)-Pdx-1-Cre mice (Pan(kras/p53) mice) were fed an AIN93M diet with or without 200 p.p.m. sulindac (n = 20/group). Kaplan-Meier survival analysis showed that average animal survival in Pan(kras/p53) mice was 143.7 ± 8.8 days, and average survival with sulindac was increased to 168.0 ± 8.8 days (P < 0.005). Histopathological analyses revealed that 90% of mice developed PDAC, 10% with metastasis to the liver and lymph nodes. With sulindac, the incidence of PDAC was reduced to 56% (P < 0.01) and only one mouse had lymph node metastasis. Immunochemical analysis showed that sulindac significantly decreased Ki-67-labeled cell proliferation and markedly reduced the expression of phosphorylated extracellular signal-regulated kinases 1 and 2 (ERK1/2), c-Raf and mitogen-activated protein kinase kinase 1 and 2. In in vitro experiments with PDAC cells from Pan(kras/p53) mice, sulindac exhibited dose-dependent inhibition of AKR1B10 activity. By silencing AKR1B10 expression through small interfering RNA or by sulindac treatment, these in vitro models showed a reduction in Kras and human DNA-J homolog 2 protein prenylation, and downregulation of phosphorylated C-raf, ERK1/2 and MEK1/2 expression. Our results demonstrate that sulindac inhibits pancreatic carcinogenesis by the inhibition of Kras protein prenylation by targeting AKR1B10.

  6. Observational study of generation conditions of substorm-associated low-frequency AKR emissions

    Directory of Open Access Journals (Sweden)

    A. Olsson

    2004-11-01

    Full Text Available It has lately been shown that low-frequency bursts of auroral kilometric radiation (AKR are nearly exclusively associated with substorm expansion phases. Here we study low-frequency AKR using Polar PWI and Interball POLRAD instruments to constrain its possible generation mechanisms. We find that there are more low-frequency AKR emission events during wintertime and equinoxes than during summertime. The dot-AKR emission radial distance range coincides well with the region where the deepest density cavities are seen statistically during Kp>2. We suggest that the dot-AKR emissions originate in the deepest density cavities during substorm onsets. The mechanism for generating dot-AKR is possibly strong Alfvén waves entering the cavity from the magnetosphere and changing their character to more inertial, which causes the Alfvén wave associated parallel electric field to increase. This field may locally accelerate electrons inside the cavity enough to produce low-frequency AKR emission. We use Interball IESP low-frequency wave data to verify that in about half of the cases the dot-AKR is accompanied by low-frequency wave activity containing a magnetic component, i.e. probably inertial Alfvén waves. Because of the observational geometry, this result is consistent with the idea that inertial Alfvén waves might always be present in the source region when dot-AKR is generated. The paper illustrates once more the importance of radio emissions as a powerful remote diagnostic tool of auroral processes, which is not only relevant for the Earth's magnetosphere but may be relevant in the future in studying extrasolar planets.

  7. New training reactor at Dresden Technical University

    International Nuclear Information System (INIS)

    Hansen, W.; Knorr, J.; Wolf, T.

    2006-01-01

    A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

  8. Leukemia prevention and long-term survival of AKR mice transplanted with MHC-matched or MHC-mismatched bone marrow

    International Nuclear Information System (INIS)

    Longley, R.E.; Good, R.A.

    1986-01-01

    The current studies were designed to evaluate the effectiveness of marrow transplantation within and outside the major histocompatibility complex (MHC) on the long-term survival and occurrence of spontaneous leukemia in AKR mice. AKR mice, which were lethally irradiated and received MHC-matched marrow from CBA/J mice (CBA----AKR), never developed leukemia and were alive and remained healthy for up to 280 days post-transplant. These long-term surviving chimeras possessed substantial immune vigor when both cell-mediated and humoral responses were tested. Lethally irradiated AKR mice, which had received MHC-mismatched marrow (anti-Thy-1.2 treated or nontreated) from C57BL/6J mice (B6----AKR), never developed leukemia and survived up to 170 days post-transplant. However, both groups of these chimeras began dying 180 to 270 days post-transplant due to a disease process which could not be readily identified. Histological analysis of B6----AKR chimeras revealed severe lymphoid cell depletion in thymus and spleen; however, none of these chimeras exhibited classical features of acute graft versus host disease. Concanavalin A mitogenesis, primary antibody responses to sheep red blood cells and the production of interleukin 2 (IL-2) were suppressed in B6----AKR chimeras. IL-2 treatment of B6----AKR chimeras was shown to partially correct these deficiencies without stimulating mixed lymphocyte responsiveness to donor or host lymphocytes. These studies indicate that the use of MHC-mismatched marrow for the prevention of spontaneous AKR leukemia may rely on augmentative IL-2 therapy for complete immune reconstitution of leukemia-free chimeras

  9. Modulation of AKR1C2 by curcumin decreases testosterone production in prostate cancer.

    Science.gov (United States)

    Ide, Hisamitsu; Lu, Yan; Noguchi, Takahiro; Muto, Satoru; Okada, Hiroshi; Kawato, Suguru; Horie, Shigeo

    2018-04-01

    Intratumoral androgen biosynthesis has been recognized as an essential factor of castration-resistant prostate cancer. The present study investigated the effects of curcumin on the inhibition of intracrine androgen synthesis in prostate cancer. Human prostate cancer cell lines, LNCaP and 22Rv1 cells were incubated with or without curcumin after which cell proliferation was measured at 0, 24, 48 and 72 hours, respectively. Prostate tissues from the transgenic adenocarcinoma of the mouse prostate (TRAMP) model were obtained after 1-month oral administration of 200 mg/kg/d curcumin. Testosterone and dihydrotestosterone concentrations in LNCaP prostate cancer cells were determined through LC-MS/MS assay. Curcumin inhibited cell proliferation and induced apoptosis of prostate cancer cells in a dose-dependent manner. Curcumin decreased the expression of steroidogenic acute regulatory proteins, CYP11A1 and HSD3B2 in prostate cancer cell lines, supporting the decrease of testosterone production. After 1-month oral administration of curcumin, Aldo-Keto reductase 1C2 (AKR1C2) expression was elevated. Simultaneously, decreased testosterone levels in the prostate tissues were observed in the TRAMP mice. Meanwhile, curcumin treatments considerably increased the expression of AKR1C2 in prostate cancer cell lines, supporting the decrease of dihydrotestosterone. Taken together, these results suggest that curcumin's natural bioactive compounds could have potent anticancer properties due to suppression of androgen production, and this could have therapeutic effects on prostate cancer. © 2018 The Authors. Cancer Science published by John Wiley & Sons Australia, Ltd on behalf of Japanese Cancer Association.

  10. Long-term In Vitro Treatment of Human Glioblastoma Cells with Temozolomide Increases Resistance In Vivo through Up-regulation of GLUT Transporter and Aldo-Keto Reductase Enzyme AKR1C Expression

    Directory of Open Access Journals (Sweden)

    Benjamin Le Calvé

    2010-09-01

    Full Text Available Glioblastoma (GBM is the most frequent malignant glioma. Treatment of GBM patients is multimodal with maximum surgical resection, followed by concurrent radiation and chemotherapy with the alkylating drug temozolomide (TMZ. The present study aims to identify genes implicated in the acquired resistance of two human GBM cells of astrocytic origin, T98G and U373, to TMZ. Resistance to TMZ was induced by culturing these cells in vitro for months with incremental TMZ concentrations up to 1 mM. Only partial resistance to TMZ has been achieved and was demonstrated in vivo in immunocompromised mice bearing orthotopic U373 and T98G xenografts. Our data show that long-term treatment of human astroglioma cells with TMZ induces increased expression of facilitative glucose transporter/solute carrier GLUT/SLC2A family members, mainly GLUT-3, and of the AKR1C family of proteins. The latter proteins are phase 1 drug-metabolizing enzymes involved in the maintenance of steroid homeostasis, prostaglandin metabolism, and metabolic activation of polycyclic aromatic hydrocarbons. GLUT-3 has been previously suggested to exert roles in GBM neovascularization processes, and TMZ was found to exert antiangiogenic effects in experimental gliomas. AKR1C1 was previously shown to be associated with oncogenic potential, with proproliferative effects similar to AKR1C3 in the latter case. Both AKR1C1 and AKR1C2 proteins are involved in cancer pro-proliferative cell chemoresistance. Selective targeting of GLUT-3 in GBM and/or AKR1C proteins (by means of jasmonates, for example could thus delay the acquisition of resistance to TMZ of astroglioma cells in the context of prolonged treatment with this drug.

  11. New control and safety rod unit for the training reactor of the Dresden Technical University

    International Nuclear Information System (INIS)

    Adam, E.; Schab, J.; Knorr, J.

    1983-01-01

    The extension of the experimental training of students at the training reactor AKR of the Dresden Technical University requires the reconstruction of the reactor with a new control and safety rod unit. The specific conditions at the AKR led to a new variant. Results of preliminary experiments, design and mode of operation of the first unit as well as hitherto gained operation experiences are presented. (author)

  12. Combined Transcriptomic Analysis Revealed AKR1B10 Played an Important Role in Psoriasis through the Dysregulated Lipid Pathway and Overproliferation of Keratinocyte

    Directory of Open Access Journals (Sweden)

    Yunlu Gao

    2017-01-01

    Full Text Available RNA-seq has enabled in-depth analysis of the pathogenesis of psoriasis on the transcriptomic level, and many biomarkers have been discovered to be related to the immune response, lipid metabolism, and keratinocyte proliferation. However, few studies have combined analysis from various datasets. In this study, we integrated different psoriasis RNA-seq datasets to reveal the pathogenesis of psoriasis through the analysis of differentially expressed genes (DEGs, pathway analysis, and functional annotation. The revealed biomarkers were further validated through proliferation phenotypes. The results showed that DEGs were functionally related to lipid metabolism and keratinocyte differentiation dysregulation. The results also showed new biomarkers, such as AKR1B10 and PLA2G gene families, as well as pathways that include the PPAR signaling pathway, cytokine-cytokine receptor interaction, alpha-linoleic acid metabolism, and glycosphingolipid biosynthesis. Using siRNA knockdown assays, we further validated the role that the AKR1B10 gene plays in proliferation. Our study demonstrated not only the dysfunction of the AKR1B10 gene in lipid metabolizing but also its important role in the overproliferation and migration of keratinocyte, which provided evidence for further therapeutic uses for psoriasis.

  13. Integration of HPV6 and downregulation of AKR1C3 expression mark malignant transformation in a patient with juvenile-onset laryngeal papillomatosis.

    Directory of Open Access Journals (Sweden)

    Christian Ulrich Huebbers

    Full Text Available Juvenile-onset recurrent respiratory papillomatosis (RRP is associated with low risk human papillomavirus (HPV types 6 and 11. Malignant transformation has been reported solely for HPV11-associated RRP in 2-4% of all RRP-cases, but not for HPV6. The molecular mechanisms in the carcinogenesis of low risk HPV-associated cancers are to date unknown. We report of a female patient, who presented with a laryngeal carcinoma at the age of 24 years. She had a history of juvenile-onset RRP with an onset at the age of three and subsequently several hundred surgical interventions due to multiple recurrences of RRP. Polymerase chain reaction (PCR or bead-based hybridization followed by direct sequencing identified HPV6 in tissue sections of previous papilloma and the carcinoma. P16(INK4A, p53 and pRb immunostainings were negative in all lesions. HPV6 specific fluorescence in situ hybridization (FISH revealed nuclear staining suggesting episomal virus in the papilloma and a single integration site in the carcinoma. Integration-specific amplification of papillomavirus oncogene transcripts PCR (APOT-PCR showed integration in the aldo-keto reductase 1C3 gene (AKR1C3 on chromosome 10p15.1. ArrayCGH detected loss of the other gene copy as part of a deletion at 10p14-p15.2. Western blot analysis and immunohistochemistry of the protein AKR1C3 showed a marked reduction of its expression in the carcinoma. In conclusion, we identified a novel molecular mechanism underlying a first case of HPV6-associated laryngeal carcinoma in juvenile-onset RRP, i.e. that HPV6 integration in the AKR1C3 gene resulted in loss of its expression. Alterations of AKR1C gene expression have previously been implicated in the tumorigenesis of other (HPV-related malignancies.

  14. Leukemia in AKR mice. III. Size distribution of suppressor T-cells in AKR leukemia and neonatal mice

    International Nuclear Information System (INIS)

    Mulder, A.M.; Durdik, J.M.; Toth, P.; Golub, E.S.

    1978-01-01

    Suppression of in vitro antibody forming potential of normal cells by leukemic cells of AKR and normal neonatal mice have many similarities. In both cases the suppression is by cell contact rather than by the elaboration of soluble suppressive factors and the suppression is sensitive to both x-irradiation and mitomycin C treatment. When the size distribution of suppressing cells in thymus and spleen were compared by velocity sedimentation, both leukemic and neonatal suppressing cells had similar size distribution in each organ. Both large and small cells in the thymus suppress but only large cells (sedimentation velocity > 3.5 mm/hr) in the spleen are able to suppress. Leukemic cells in lymph node have a splenic size distribution, viz., only large cells suppress. Both large and small cells of a subcutaneously growing long passage AKR lymphoma are able to suppress. While large cells contain the bulk of cells actively incorporating tritiated thymidine and thus probably in cycle, small but significant amounts of incorporation in small suppressing cells is also seen

  15. Beneficial metabolic effects of CB1R anti-sense oligonucleotide treatment in diet-induced obese AKR/J mice.

    Directory of Open Access Journals (Sweden)

    Yuting Tang

    Full Text Available An increasing amount of evidence supports pleiotropic metabolic roles of the cannibinoid-1 receptor (CB1R in peripheral tissues such as adipose, liver, skeletal muscle and pancreas. To further understand the metabolic consequences of specific blockade of CB1R function in peripheral tissues, we performed a 10-week-study with an anti-sense oligonucleotide directed against the CB1R in diet-induced obese (DIO AKR/J mice. DIO AKR/J mice were treated with CB1R ASO Isis-414930 (6.25, 12.5 and 25 mg/kg/week or control ASO Isis-141923 (25 mg/kg/week via intraperitoneal injection for 10 weeks. At the end of the treatment, CB1R mRNA from the 25 mg/kg/week CB1R ASO group in the epididymal fat and kidney was decreased by 81% and 63%, respectively. Body weight gain was decreased in a dose-dependent fashion, significantly different in the 25 mg/kg/week CB1R ASO group (46.1±1.0 g vs veh, 51.2±0.9 g, p<0.05. Body fat mass was reduced in parallel with attenuated body weight gain. CB1R ASO treatment led to decreased fed glucose level (at week 8, 25 mg/kg/week group, 145±4 mg/dL vs veh, 195±10 mg/dL, p<0.05. Moreover, CB1R ASO treatment dose-dependently improved glucose excursion during an oral glucose tolerance test, whereas control ASO exerted no effect. Liver steatosis was also decreased upon CB1R ASO treatment. At the end of the study, plasma insulin and leptin levels were significantly reduced by 25 mg/kg/week CB1R ASO treatment. SREBP1 mRNA expression was decreased in both epididymal fat and liver. G6PC and fatty acid translocase/CD36 mRNA levels were also reduced in the liver. In summary, CB1R ASO treatment in DIO AKR/J mice led to improved insulin sensitivity and glucose homeostasis. The beneficial effects of CB1R ASO treatment strongly support the notion that selective inhibition of the peripheral CB1R, without blockade of central CB1R, may serve as an effective approach for treating type II diabetes, obesity and the metabolic syndrome.

  16. Identification of the common radiation-sensitive and glucose metabolism-related expressed genes in the thymus of ICR and AKR/J mice

    International Nuclear Information System (INIS)

    Bong, Jin Jong; Kang, Yumi; Choi, Suk Cjul; Choi, Moo Hyun; Choi, Seung Jin; Kim, Hee Sun

    2011-01-01

    Our goal was to identify the common radiation-sensitive expressed genes in the thymus of ICR and AKR/J mice on 100 days after irradiation. Thus, we performed microarray analysis for thymus of ICR and AKR/J mice, respectively. We categorized differential expressed genes by the analysis of DAVID Bioinformatics Resources v 6.7 and GeneSpring GX 11.5.1 and validated gene expression patterns by QPCR analysis. Our result demonstrated that radiation-sensitive expressed genes and signaling pathways in the thymus of irradiated ICR and AKR/J mice.

  17. Development of donor-derived thymic lymphomas after allogeneic bone marrow transplantation in AKR/J mice

    International Nuclear Information System (INIS)

    Yasumizu, R.; Hiai, H.; Sugiura, K.

    1988-01-01

    The transplantation of bone marrow cells from BALB/c (but not C57BL/6 and C3H/HeN) mice was observed to lead to the development of thymic lymphomas (leukemias) in AKR/J mice. Two leukemic cell lines, CAK1.3 and CAK4.4, were established from the primary culture of two thymic lymphoma, and surface phenotypes of these cell lines found to be H-2d and Thy-1.2+, indicating that these lymphoma cells are derived from BALB/c donor bone marrow cells. Further analyses of surface markers revealed that CAK1.3 is L3T4+ Lyt2+ IL2R-, whereas CAK4.4 is L3T4- Lyt2- IL2R+. Both CAK1.3 and CAK4.4 were transplantable into BALB/c but not AKR/J mice, further indicating that these cells are of BALB/c bone marrow donor origin. The cells were found to produce XC+-ecotropic viruses, but xenotropic and mink cell focus-forming viruses were undetectable. Inasmuch as thymic lymphomas are derived from bone marrow cells of leukemia-resistant BALB/c strain of mice under the allogeneic environment of leukemia-prone AKR/J mice, this animal model may serve as a useful tool not only for the analysis of leukemic relapse after bone marrow transplantation but also for elucidation of the mechanism of leukemogenesis

  18. Expression of antigens coded in murine leukemia viruses on thymocytes of allogeneic donor origin in AKR mice following syngeneic or allogeneic bone marrow transplantation

    International Nuclear Information System (INIS)

    Wustrow, T.P.; Good, R.A.

    1985-01-01

    Removal of T-lymphocytes from marrow inoculum with monoclonal antibody plus complement permitted establishment of long-lived allogeneic chimeras between C57BL/6 and AKR/J mice. Development of leukemia was prevented for 15 mo. Protection from leukemia occurred with both young (4 wk) and older (4 mo) recipients. AKR mice reconstituted with syngeneic marrow or control AKR mice all developed leukemia-lymphoma before 1 yr of age. During spontaneous lymphomagenesis in AKR mice, amplified expression of gag or env gene-coded virus antigens on the surface of thymocytes preceded leukemia development and evidence for amplification of other virus genes. These changes generally appeared before 6 mo. Similar viral gene expression and viral gene amplification occurred in the thymus and spleen cells of leukemia-resistant chimeric mice. Using monoclonal antibodies to Mr 70,000 glycoprotein epitopes characteristic of ecotropic, xenotropic, or dualtropic viruses, antigens marking each virus form were found on thymocytes of allogeneic 4-wk and 4-mo chimeras as well as on the cells of AKR mice and of AKR mice reconstituted with syngeneic marrow. Flow cytometric analysis showed amplification of the virus genes in mice protected from leukemia-lymphoma by allogeneic bone marrow transplantation from leukemia-resistant mice. Allogeneic chimeras and syngeneically transplanted mice both showed evidence of accelerated viremia and of recombinant virus formation. The findings suggest that an event essential to leukemogenesis which occurs within the AKR lymphoid cells or their environment is lacking in the allogeneic chimeras. The nature of this influence of a resistance gene or genes introduced into AKR mice by allogeneic bone marrow transplantation deserves further study

  19. Ray tracing of auroral Z mode radiation, AKR and auroral hiss

    International Nuclear Information System (INIS)

    Horne, R.B.; Jones, D.; Kimura, I.; Sawada, A.

    1990-01-01

    While observed frequency bandwidths of auroral Z mode radiation cannot be directly accounted for in terms of direct cyclotron maser instability generation, ray tracing in a hot plasma indicates that if the radiation near a plasma frequency lower than the gyrofrequency, the observed bandwidths are explainable in terms of upward propagation away from the earth. An auroral Z-mode generation mechanism is proposed involving mode conversion from O-mode auroral kilometric radiation (AKR) at the plasma frequency, as well as mode conversion from upgoing auroral hiss. Ray tracings in the O mode identify a possible AKR source region along L = 8.55. 11 refs

  20. A novel aldo-keto reductase from Jatropha curcas L. (JcAKR) plays a crucial role in the detoxification of methylglyoxal, a potent electrophile.

    Science.gov (United States)

    Mudalkar, Shalini; Sreeharsha, Rachapudi Venkata; Reddy, Attipalli Ramachandra

    2016-05-20

    Abiotic stress leads to the generation of reactive oxygen species (ROS) which further results in the production of reactive carbonyls (RCs) including methylglyoxal (MG). MG, an α, β-dicarbonyl aldehyde, is highly toxic to plants and the mechanism behind its detoxification is not well understood. Aldo-keto reductases (AKRs) play a role in detoxification of reactive aldehydes and ketones. In the present study, we cloned and characterised a putative AKR from Jatropha curcas (JcAKR). Phylogenetically, it forms a small clade with AKRs of Glycine max and Rauwolfia serpentina. JcAKR was heterologously expressed in Escherichia coli BL-21(DE3) cells and the identity of the purified protein was confirmed through MALDI-TOF analysis. The recombinant protein had high enzyme activity and catalytic efficiency in assays containing MG as the substrate. Protein modelling and docking studies revealed MG was efficiently bound to JcAKR. Under progressive drought and salinity stress, the enzyme and transcript levels of JcAKR were higher in leaves compared to roots. Further, the bacterial and yeast cells expressing JcAKR showed more tolerance towards PEG (5%), NaCl (200mM) and MG (5mM) treatments compared to controls. In conclusion, our results project JcAKR as a possible and potential target in crop improvement for abiotic stress tolerance. Copyright © 2016 Elsevier GmbH. All rights reserved.

  1. Stimulation of anti-tumor effect by low-dose irradiation. Pt. 2. The prolongation of life span in AKR mice

    International Nuclear Information System (INIS)

    Ishii, Keiichiro; Misonoh, Jun; Hosoi, Yoshio; Ono, Tetsuya; Sakamoto, Kiyohiko.

    1994-01-01

    To elucidate the antileukemic effect of low-dose X-irradiation, we studied the influence of periodical low-dose X-irradiation on survival and tumor incidence of thymus using AKR mice. The findings of the experiments were as follows; (1) The median survival time of control AKR mice was 283±3 days. It of irradiation group of 15 cGy/week and 30cGy was 309±14 days and 316±10 days respectively. The life span was significantly prolonged (p < 0.05 and p < 0.01 respectively by Wilcoxon test) by periodical low-dose X-irradiation in term of breeding. (2) The incidence of thymus tumor which is observed remarkably in control AKR mice was 48.8%. It of irradiation group of 15 cGy/week and 30 cGy/week was 40% and 20% respectively. Inversely, the non-tumor incidence of tymus in control AKR mice was 19.5%. It of irradiation group of 15 cGy/week and 30 cGy/week was 32.5% and 51.4% respectively. The thymic tumor incidence was significantly decreased (p < 0.01 by chi-square test) in irradiation group of 30 cGy/week. (3) The incidence of thymic lymphoma as a death cause in control AKR mice was 80.4%. It of irradiation group of 15 cGy/week and 30cGy/week was 67.5% and 48.6% respectively. The incidence of thymic lymphoma was significantly decreased (p < 0.05 by chi-square test) in irradiation group of 30 cGy/week. (author)

  2. Suppressive effects of Lactobacillus casei cells, a bacterial immunostimulant, on the incidence of spontaneous thymic lymphoma in AKR mice.

    Science.gov (United States)

    Watanabe, T

    1996-06-01

    The mean survival age of female AKR/J mice was significantly prolonged, the enlargement of thymus was markedly suppressed, and the proliferation of ecotropic and recombinant murine leukemia viruses (MuLV) was markedly inhibited when 8-week-old female AKR/J mice were injected intraperitoneally (i.p.) with heat-killed Lactobacillus casei cells twice weekly for 8 weeks. In contrast, such actions of heat-killed L. casei cells were not seen in 20-week-old female AKR/J mice. The leukemogenic activity of the cell-free extract of thymus from adult female AKR/J mice in newborn female AKR/J mice was drastically reduced by i.p. treatment with heat-killed L. casei cells. The difference in adjuvant effectiveness of heat-killed L. casei cells on 8- and 20-week-old animals may be dependent on the difference in the enhancing activity of the cell-mediated immune systems between the groups induced by heat-killed L. casei cells, and, as a result, on the difference in the degree of proliferation of ecotropic and recombinant MuLV in thymus, which consequently causes thymic lymphoma.

  3. Estimation of tail reconnection lines by AKR onsets and plasmoid entries observed with GEOTAIL spacecraft

    International Nuclear Information System (INIS)

    Murata, Takeshi; Matsumoto, Hiroshi; Kojima, Hirotsugu

    1995-01-01

    We estimate the location of the reconnection line and plasmoid size in the geomagnetic tail using data from the Plasma Wave Instrument onboard the GEOTAIL spacecraft. We first compare AKR onset events with high energy particle observations at geosynchronous orbit. We determine the plasmoid ejection (re-connection) time by the AKR enhancement only when it corrresponds to energetic particle enhancement within five minutes. The traveling time of the plasmoid from the X-line to the spacecraft is calculated by the difference in time of the AKR onset and that of the plasmoid encounter with GEOTAIL. Assuming the plasmoid propagates with the Alfven velocity in the tail lobe as MHD simulations predict, we estimate the location of the reconnection line in 11 events. The results show that the most probable location of the plasmoid edge is distributed around Χ = -60 R E in the GSE coordinates. The estimated size of the plasmoids ranges from 10 to 50 R E in the χ direction. If we apply this result to the alternative plasmoid model in which the evolution of the tearing instability causes the generation of plasmoids, the X-line should be approximately at χ = -35 R E . 15 refs., 3 figs., 1 tab

  4. Crystal Structure of Perakine Reductase, Founding Member of a Novel Aldo-Keto Reductase (AKR) Subfamily That Undergoes Unique Conformational Changes during NADPH Binding*

    Science.gov (United States)

    Sun, Lianli; Chen, Yixin; Rajendran, Chitra; Mueller, Uwe; Panjikar, Santosh; Wang, Meitian; Mindnich, Rebekka; Rosenthal, Cindy; Penning, Trevor M.; Stöckigt, Joachim

    2012-01-01

    Perakine reductase (PR) catalyzes the NADPH-dependent reduction of the aldehyde perakine to yield the alcohol raucaffrinoline in the biosynthetic pathway of ajmaline in Rauvolfia, a key step in indole alkaloid biosynthesis. Sequence alignment shows that PR is the founder of the new AKR13D subfamily and is designated AKR13D1. The x-ray structure of methylated His6-PR was solved to 2.31 Å. However, the active site of PR was blocked by the connected parts of the neighbor symmetric molecule in the crystal. To break the interactions and obtain the enzyme-ligand complexes, the A213W mutant was generated. The atomic structure of His6-PR-A213W complex with NADPH was determined at 1.77 Å. Overall, PR folds in an unusual α8/β6 barrel that has not been observed in any other AKR protein to date. NADPH binds in an extended pocket, but the nicotinamide riboside moiety is disordered. Upon NADPH binding, dramatic conformational changes and movements were observed: two additional β-strands in the C terminus become ordered to form one α-helix, and a movement of up to 24 Å occurs. This conformational change creates a large space that allows the binding of substrates of variable size for PR and enhances the enzyme activity; as a result cooperative kinetics are observed as NADPH is varied. As the founding member of the new AKR13D subfamily, PR also provides a structural template and model of cofactor binding for the AKR13 family. PMID:22334702

  5. Structural variations in the H-2 genes of AKR lymphomas.

    NARCIS (Netherlands)

    K. Hui; L. Minamide; N. Prandoni; H. Festenstein; F.G. Grosveld (Frank)

    1986-01-01

    textabstractK36.16 is an AKR H-2k thymoma which expresses an aberrant H-2Dd-like allospecificity, does not have a detectable amount of the H-2Kk syngeneic antigen and grows very easily in syngeneic mice. By DNA-mediated gene transfer experiments, we were able to obtain transformed clones which do

  6. Investigations of the reactivity temperature coefficient of the Dresden Technical University training and research reactor

    International Nuclear Information System (INIS)

    Adam, E.; Knorr, J.

    1982-01-01

    Approximate formulas are derived for determining the temperature coefficient of reactivity of the training and research reactor (AKR) of the Dresden Technical University. Values calculated on the basis of these approximations show good agreement with experimentally obtained results, thus confirming the applicability of the formulas to simple systems

  7. The role of KURT and A-KRS in the development of generic safety cases in Korea

    International Nuclear Information System (INIS)

    Jeong, Jong-Tae; Choi, Heui-Joo; Koh, Yong-Kwon; Kim, Geon-Young; Kim, Kyung-Su

    2014-01-01

    According to the draft guidelines for a deep geological disposal system for high-level wastes in Korea, the total annual risk for the average person resulting from the radiation exposure should not exceed 1.0 x 10 -6 /yr and the expected radiation exposure to the average person for each scenario should not exceed 10 mSv/yr (NSSC, 2012). Regulatory compliance should be supported by multiple lines of reasoning such as probabilistic analysis of exposure dose and risk, uncertainty analysis, natural analogue, complementary safety indicators such as radionuclide concentration and release rates, secure of defence-in-depth. The integrated safety assessment should also be made and updated consistently based on up-to-date data and information for each stage of deep geological disposal of high-level waste (HLW) such as basic studies, site characterisation, design, construction, operation, closure, environmental monitoring after closure and so on. These are the bases for the development of a safety case in Korea. The Korea Atomic Energy Research Institute (KAERI) is now developing generic safety cases based on the Advanced Korean Reference Repository System (A-KRS) and the KAERI Underground Research Tunnel (KURT). The A-KRS is a geological disposal system for radioactive wastes from the pyro-processing of PWR spent nuclear fuels in Korea. KURT is a small-scale underground research laboratory constructed and operated to assess the feasibility, safety, appropriateness and stability of the disposal concept by making various in situ tests and experiments. In this paper, the concept and long-term safety assessment of the A-KRS design, the role of the KURT in the development of safety cases, and the development of complex scenarios to support the development of generic safety cases in Korea are described. (authors)

  8. Thy-1+ dendritic cells in murine epidermis are bone marrow-derived

    International Nuclear Information System (INIS)

    Breathnach, S.M.; Katz, S.I.

    1984-01-01

    Thy-1+, Ly-5+ dendritic cells have recently been described as a resident cell population in murine epidermis, but their ontogeny and function are unknown. The origin and turnover of epidermal Thy-1+ cells utilizing chimeric mice were investigated. Lethally x-irradiated AKR/J (Thy-1.1+) and AKR/Cum (Thy-1.2+) mice were reconstituted with allogeneic bone marrow cells with or without thymocytes from congenic AKR/Cum or AKR/J mice, respectively. The density of residual indigenous Thy-1.1+ cells in AKR/J chimeras and Thy-1.2+ cells in AKR/Cum chimeras was substantially reduced following x-irradiation, as determined by immunofluorescence staining of epidermal sheets. Epidermal repopulation by allogeneic Thy-1+ dendritic epidermal cells was first observed at 5 weeks in AKR/J chimeras and at 7 weeks in AKR/Cum chimeras and progressed slowly. Repopulation was not enhanced by increasing the number of allogeneic bone marrow cells injected from 2 X 10(7) to 10(8) cells or by the addition of 8 X 10(7) allogeneic thymocytes to the donor inoculate. Epidermal repopulation by allogeneic Thy-1.2+ cells was not seen in AKR/J mice reconstituted with syngeneic bone marrow cells and allogeneic Thy-1.2+ AKR/Cum thymocytes. Taken together, these results indicate that Thy-1+ dendritic epidermal cells are derived from the bone marrow and suggest that they are not related to conventional peripheral T-lymphocytes

  9. Cytogenetic analysis of X-ray induced chromosome aberrations in spontaneous leukaemic AKR mice

    International Nuclear Information System (INIS)

    Szollar, J.

    1975-01-01

    The increased frequency of numerical and structural chromosomal aberrations in spontaneously leukaemic AKR mice, compared with the values of healthy control CBA/H-T 6 T 6 mice, induced by X-irradiation, might be connected with the predisposition to malignant growth, probably indirectly helping the virus activation, or acting together with the immune deficiency, by creating a weaker system that is more sensitive to carcinogenic agents

  10. Detection of calmodulin binding protein at 170 KDA in BALB, AKR, DON and chicken granulosa cells

    International Nuclear Information System (INIS)

    Selinfreund, R.; Lin, P.H.; Marrone, B.; Wharton, W.

    1987-01-01

    Calmodulin (CAM) has been shown to bind to the epidermal growth factor (EGF) receptor (170 kDa) and is phosphorylated in a EGF dependent manner in the A431 human epidermoid carcinoma cells. In the present study, they report 125 I-CAM binding to a 170 kDa protein detected in cell membrane vesicles of Balb/3T3, AKR, DON and chicken granulosa cells. Purified plasma membranes from these cells were resolved via electrophoresis (without heat denaturation) and electroblotted onto nictrocellulose paper. Upon hybridizing against 125 I-CAM, a distinct autoradiographic band occurred at 170 kDa for all the cells lines under study. The binding of CAM is specific and can be displaced with the addition of excess unlabeled CAM. The result suggest that 125 I-CAM may bind to the 170 kDa EGF receptor in BALB, AKR, DON and chicken granulosa cells

  11. Anthracycline resistance mediated by reductive metabolism in cancer cells: The role of aldo-keto reductase 1C3

    International Nuclear Information System (INIS)

    Hofman, Jakub; Malcekova, Beata; Skarka, Adam; Novotna, Eva; Wsol, Vladimir

    2014-01-01

    Pharmacokinetic drug resistance is a serious obstacle that emerges during cancer chemotherapy. In this study, we investigated the possible role of aldo-keto reductase 1C3 (AKR1C3) in the resistance of cancer cells to anthracyclines. First, the reducing activity of AKR1C3 toward anthracyclines was tested using incubations with a purified recombinant enzyme. Furthermore, the intracellular reduction of daunorubicin and idarubicin was examined by employing the transfection of A549, HeLa, MCF7 and HCT 116 cancer cells with an AKR1C3 encoding vector. To investigate the participation of AKR1C3 in anthracycline resistance, we conducted MTT cytotoxicity assays with these cells, and observed that AKR1C3 significantly contributes to the resistance of cancer cells to daunorubicin and idarubicin, whereas this resistance was reversible by the simultaneous administration of 2′-hydroxyflavanone, a specific AKR1C3 inhibitor. In the final part of our work, we tracked the changes in AKR1C3 expression after anthracycline exposure. Interestingly, a reciprocal correlation between the extent of induction and endogenous levels of AKR1C3 was recorded in particular cell lines. Therefore, we suggest that the induction of AKR1C3 following exposure to daunorubicin and idarubicin, which seems to be dependent on endogenous AKR1C3 expression, eventually might potentiate an intrinsic resistance given by the normal expression of AKR1C3. In conclusion, our data suggest a substantial impact of AKR1C3 on the metabolism of daunorubicin and idarubicin, which affects their pharmacokinetic and pharmacodynamic behavior. In addition, we demonstrate that the reduction of daunorubicin and idarubicin, which is catalyzed by AKR1C3, contributes to the resistance of cancer cells to anthracycline treatment. - Highlights: • Metabolism of anthracyclines by AKR1C3 was studied at enzyme and cellular levels. • Anthracycline resistance mediated by AKR1C3 was demonstrated in cancer cells. • Induction of AKR1C3

  12. Transforming growth factor type β can act as a potent competence factor for AKR-2B cells

    International Nuclear Information System (INIS)

    Goustin, A.S.; Nuttall, G.A.; Leof, E.B.; Ranganathan, G.; Moses, H.L.

    1987-01-01

    Transforming growth factor type β (TGFβ) is a pleiotropic regulator of cell growth with specific high-affinity cell-surface receptors on a large number of cells; its mechanism of action, however, is poorly defined. In this report, the authors utilized the mouse fibroblast line AKR-2B to explore the question of the temporal requirements during the cell cycle in regard to both the growth inhibitory and the growth stimulatory action of TGFβ. The results indicate that AKR-2B cells are most sensitive to the inhibitory action of TGFβ during early to mid-G 1 . In addition, TGFβ need be present only briefly in order to exert its inhibitory effect on EGF-induced DNA synthesis. Likewise, the stimulatory effect of TGFβ in the absence of EGF requires only an equally brief exposure to TGFβ. Use of homogeneous 125 I-labeled TGFβ in a cell-binding assay demonstrates that TGFβ bound to cell-surface receptors can readily exchange into the culture medium, helping to rule out the possibility that persistent receptor-bound TGFβ is the source of a continuous stimulus. The data indicate that TGFβ exposure induces a stable state in the cell similar to but distinct from the state of competence induced by platelet-derived growth factor (PDGF)

  13. Polarization measurements of auroral kilometric radiation by Dynamics Explorer-1

    International Nuclear Information System (INIS)

    Shawhan, S.D.; Gurnett, D.A.

    1982-01-01

    The plasma wave instrument (PWI) on the Dynamics Explorer-1 has been used to measure polarization of auroral kilometric radiation (AKR) at frequencies of 50 to 400 kHz in both the northern and the southern nightside auroral regions at altitudes of 1 to 3 R/sub E/ above the AKR source regions. The AKR polarization sense is found to be the same as the right hand polarized auroral hiss found in the frequency range of 0.8 to 6.4 kHz. Consequently, these unambiguous direct polarization measurements of AKR lead to the conclusion that AKR escapes the magnetosphere in the R-X mode. Since DE-1 is close to the source region, it can be inferred that AKR is generated predominately in the R-X mode

  14. The Role of Human Aldo-Keto Reductases (AKRs in the Metabolic Activation and Detoxication of Polycyclic Aromatic Hydrocarbons: Interconversion of PAH-catechols and PAH o-Quinones

    Directory of Open Access Journals (Sweden)

    Li eZhang

    2012-11-01

    Full Text Available Polycyclic aromatic hydrocarbons (PAH are ubiquitous environmental pollutants. They are procarcinogens requiring metabolic activation to elicit their deleterious effects. Aldo-keto reductases (AKR catalyze the oxidation of proximate carcinogenic PAH trans-dihydrodiols to yield electrophilic and redox-active PAH o-quiniones. AKRs are also found to be capable of reducing PAH o-quinones to form PAH catechols. The interconversion of o-quinones and catechols results in the redox cycling of PAH o-quinones to give rise to the generation of reactive oxygen species and subsequent oxidative DNA damage. On the other hand, PAH catechols can be intercepted through phase II metabolism by which PAH o-quinones could be detoxified and eliminated. The aim of the present review is to summarize the role of human AKRs in the metabolic activation/detoxication of PAH and the relevance of phase II conjugation reactions to human lung carcinogenesis.

  15. Investigation of intermittent magnetic flux in the auroral zones with kilometer radiation (AKR)

    International Nuclear Information System (INIS)

    Liu, S.Q.; Li, X.Q.

    2001-01-01

    On the basis of the nonlinear equations for self-generated magnetic fields, it is numerically shown that the magnetic fields self-generated are instable and may collapse, resulting in spatially highly intermittent flux fragment. Numerical results show that the enhanced magnetic flux has a strength about up to 10 -2 Gauss in range about around 250-350 km in auroral zones with kilometric radiation (AKR), which correspond to estimated values in both the strength and characteristic scale by Mckean et al. [J. Geophys. Res. [Oceans] 96, 21055 (1991)

  16. Metabolism of trans, trans-muconaldehyde, a cytotoxic metabolite of benzene, in mouse liver by alcohol dehydrogenase Adh1 and aldehyde reductase AKR1A4

    International Nuclear Information System (INIS)

    Short, Duncan M.; Lyon, Robert; Watson, David G.; Barski, Oleg A.; McGarvie, Gail; Ellis, Elizabeth M.

    2006-01-01

    The reductive metabolism of trans, trans-muconaldehyde, a cytotoxic metabolite of benzene, was studied in mouse liver. Using an HPLC-based stopped assay, the primary reduced metabolite was identified as 6-hydroxy-trans, trans-2,4-hexadienal (OH/CHO) and the secondary metabolite as 1,6-dihydroxy-trans, trans-2,4-hexadiene (OH/OH). The main enzymes responsible for the highest levels of reductase activity towards trans, trans-muconaldehyde were purified from mouse liver soluble fraction first by Q-sepharose chromatography followed by either blue or red dye affinity chromatography. In mouse liver, trans, trans-muconaldehyde is predominantly reduced by an NADH-dependent enzyme, which was identified as alcohol dehydrogenase (Adh1). Kinetic constants obtained for trans, trans-muconaldehyde with the native Adh1 enzyme showed a V max of 2141 ± 500 nmol/min/mg and a K m of 11 ± 4 μM. This enzyme was inhibited by pyrazole with a K I of 3.1 ± 0.57 μM. Other fractions were found to contain muconaldehyde reductase activity independent of Adh1, and one enzyme was identified as the NADPH-dependent aldehyde reductase AKR1A4. This showed a V max of 115 nmol/min/mg and a K m of 15 ± 2 μM and was not inhibited by pyrazole

  17. Effect of low-intensity low-dose rate irradiation on the incidence and the development of spontaneous leukosis in AKR mice

    International Nuclear Information System (INIS)

    Burlakova, E.B.; Erokhin, V.N.

    2001-01-01

    Development of spontaneous leukosis in AKR mice is accelerated by irradiation with low doses of 1.2-2.4 cGy and low dose rate 0.06 cGy/day. The leukoses incidence rate increases. Deaths of the animals from leukosis occurs earlier, shortening the average and maximum life-spans of the animals. The dynamics of changes in the mass of organs of the immune systems (thymus and spleen) shows extrema. The moment of reaching the extremum correlates with the maximum rate of animals' deaths [ru

  18. Expression of human aldo-keto reductase 1C2 in cell lines of peritoneal endometriosis: potential implications in metabolism of progesterone and dydrogesterone and inhibition by progestins.

    Science.gov (United States)

    Beranič, Nataša; Brožič, Petra; Brus, Boris; Sosič, Izidor; Gobec, Stanislav; Lanišnik Rižner, Tea

    2012-05-01

    The human aldo-keto reductase AKR1C2 converts 5α-dihydrotestosterone to the less active 3α-androstanediol and has a minor 20-ketosteroid reductase activity that metabolises progesterone to 20α-hydroxyprogesterone. AKR1C2 is expressed in different peripheral tissues, but its role in uterine diseases like endometriosis has not been studied in detail. Some progestins used for treatment of endometriosis inhibit AKR1C1 and AKR1C3, with unknown effects on AKR1C2. In this study we investigated expression of AKR1C2 in the model cell lines of peritoneal endometriosis, and examined the ability of recombinant AKR1C2 to metabolise progesterone and progestin dydrogesterone, as well as its potential inhibition by progestins. AKR1C2 is expressed in epithelial and stromal endometriotic cell lines at the mRNA level. The recombinant enzyme catalyses reduction of progesterone to 20α-hydroxyprogesterone with a 10-fold lower catalytic efficiency than the major 20-ketosteroid reductase, AKR1C1. AKR1C2 also metabolises progestin dydrogesterone to its 20α-dihydrodydrogesterone, with 8.6-fold higher catalytic efficiency than 5α-dihydrotestosterone. Among the progestins that are currently used for treatment of endometriosis, dydrogesterone, medroxyprogesterone acetate and 20α-dihydrodydrogesterone act as AKR1C2 inhibitors with low μM K(i) values in vitro. Their potential in vivo effects should be further studied. Copyright © 2011 Elsevier Ltd. All rights reserved.

  19. The DHEA-sulfate depot following P450c17 inhibition supports the case for AKR1C3 inhibition in high risk localized and advanced castration resistant prostate cancer.

    Science.gov (United States)

    Tamae, Daniel; Mostaghel, Elahe; Montgomery, Bruce; Nelson, Peter S; Balk, Steven P; Kantoff, Philip W; Taplin, Mary-Ellen; Penning, Trevor M

    2015-06-05

    Prostate cancer is the second leading cause of cancer death in the United States. Treatment of localized high-risk disease and de novo metastatic disease frequently leads to relapse. These metastatic castration resistant prostate cancers (mCRPC) claim a high mortality rate, despite the extended survival afforded by the growing armamentarium of androgen deprivation, radiation and immunotherapies. Here, we review two studies of neoadjuvant treatment of high-risk localized prostate cancer prior to prostatectomy, the total androgen pathway suppression (TAPS) trial and the neoadjuvant abiraterone acetate (AA) trial. These two trials assessed the efficacy of the non-specific P450c17 inhibitor, ketoconazole and the specific P450c17 inhibitor, AA, to inhibit tissue and serum androgen levels. Furthermore, a novel and validated stable isotope dilution liquid chromatography electrospray ionization selected reaction monitoring mass spectrometry assay was used to accurately quantify adrenal and gonadal androgens in circulation during the course of these trials. The adrenal androgens, Δ(4)-androstene-3,17-dione, dehydroepiandrosterone and dehydroepiandrosterone sulfate were significantly reduced in the patients receiving ketoconazole or AA compared to those who did not. However, in both trials, a significant amount of DHEA-S (∼20 μg/dL) persists and thus may serve as a depot for intratumoral conversion to the potent androgen receptor ligands, testosterone (T) and 5α-dihydrotestosterone (DHT). The final step in conversion of Δ(4)-androstene-3,17-dione and 5α-androstanedione to T and DHT, respectively, is catalyzed by AKR1C3. We therefore present the case that in the context of the DHEA-S depot, P450c17 and AKR1C3 inhibition may be an effective combinatorial treatment strategy. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  20. Crystal structures of three classes of non-steroidal anti-inflammatory drugs in complex with aldo-keto reductase 1C3.

    Directory of Open Access Journals (Sweden)

    Jack U Flanagan

    Full Text Available Aldo-keto reductase 1C3 (AKR1C3 catalyses the NADPH dependent reduction of carbonyl groups in a number of important steroid and prostanoid molecules. The enzyme is also over-expressed in prostate and breast cancer and its expression is correlated with the aggressiveness of the disease. The steroid products of AKR1C3 catalysis are important in proliferative signalling of hormone-responsive cells, while the prostanoid products promote prostaglandin-dependent proliferative pathways. In these ways, AKR1C3 contributes to tumour development and maintenance, and suggest that inhibition of AKR1C3 activity is an attractive target for the development of new anti-cancer therapies. Non-steroidal anti-inflammatory drugs (NSAIDs are one well-known class of compounds that inhibits AKR1C3, yet crystal structures have only been determined for this enzyme with flufenamic acid, indomethacin, and closely related analogues bound. While the flufenamic acid and indomethacin structures have been used to design novel inhibitors, they provide only limited coverage of the NSAIDs that inhibit AKR1C3 and that may be used for the development of new AKR1C3 targeted drugs. To understand how other NSAIDs bind to AKR1C3, we have determined ten crystal structures of AKR1C3 complexes that cover three different classes of NSAID, N-phenylanthranilic acids (meclofenamic acid, mefenamic acid, arylpropionic acids (flurbiprofen, ibuprofen, naproxen, and indomethacin analogues (indomethacin, sulindac, zomepirac. The N-phenylanthranilic and arylpropionic acids bind to common sites including the enzyme catalytic centre and a constitutive active site pocket, with the arylpropionic acids probing the constitutive pocket more effectively. By contrast, indomethacin and the indomethacin analogues sulindac and zomepirac, display three distinctly different binding modes that explain their relative inhibition of the AKR1C family members. This new data from ten crystal structures greatly broadens

  1. Radiation field studies at the training and research reactor AKR of the Dresden Technical University

    International Nuclear Information System (INIS)

    Leuschner, A.; Reiss, U.; Pretzsch, G.

    1983-01-01

    Results of radiation field studies in the experimental channels of the training and research reactor of the Technical University of Dresden are presented. The flux densities of thermal, intermediate and fast neutrons were determined by means of activation detectors., Gamma dose rates have been measured by thermoluminescent dosimeters. The measured results show symmetry with respect to the vertical axis of the reactor and allow to draw conclusions with regard to the efficiency of the individual layers of the shield. They are an essential basis of performing irradiation experiments in the experimental channels. The results of measurements were compared with those of shielding and design calculations. Taking into account the measuring errors and the approximations used in the computational models, no unexpected deviations have been observed. Hence, the measured and calculated results can be assessed to be in good agreement. (author)

  2. Aldo-keto Reductase Family 1 B10 as a Novel Target for Breast Cancer Treatment

    Science.gov (United States)

    2010-08-01

    overexpressed in tested human breast cancer tissues and mediates acetyl-CoA carboxylase-α ( ACCA ) stability, affecting fatty acid de novo synthesis and...9703; Fax. 217-545-3227; E-mail: dcao@siumed.edu Running title: AKR1B10 as a new risk factor for breast cancer Abbreviations used: ACCA , acetyl...The effect of AKR1B10 expression in cancer tissue on patient survival was evaluated with Kaplan - Meier plots, and results showed that AKR1B10

  3. Aldo-keto reductase family 1 B10 protein detoxifies dietary and lipid-derived alpha, beta-unsaturated carbonyls at physiological levels

    International Nuclear Information System (INIS)

    Zhong, Linlin; Liu, Ziwen; Yan, Ruilan; Johnson, Stephen; Zhao, Yupei; Fang, Xiubin; Cao, Deliang

    2009-01-01

    Alpha, beta-unsaturated carbonyls are highly reactive mutagens and carcinogens to which humans are exposed on a daily basis. This study demonstrates that aldo-keto reductase family 1 member B10 (AKR1B10) is a critical protein in detoxifying dietary and lipid-derived unsaturated carbonyls. Purified AKR1B10 recombinant protein efficiently catalyzed the reduction to less toxic alcohol forms of crotonaldehyde at 0.90 μM, 4-hydroxynonenal (HNE) at 0.10 μM, trans-2-hexanal at 0.10 μM, and trans-2,4-hexadienal at 0.05 μM, the concentrations at or lower than physiological exposures. Ectopically expressed AKR1B10 in 293T cells eliminated immediately HNE at 1 (subtoxic) or 5 μM (toxic) by converting to 1,4-dihydroxynonene, protecting the cells from HNE toxicity. AKR1B10 protein also showed strong enzymatic activity toward glutathione-conjugated carbonyls. Taken together, our study results suggest that AKR1B10 specifically expressed in the intestine is physiologically important in protecting the host cell against dietary and lipid-derived cytotoxic carbonyls.

  4. Exposure to 9,10-phenanthrenequinone accelerates malignant progression of lung cancer cells through up-regulation of aldo-keto reductase 1B10

    International Nuclear Information System (INIS)

    Matsunaga, Toshiyuki; Morikawa, Yoshifumi; Haga, Mariko; Endo, Satoshi; Soda, Midori; Yamamura, Keiko; El-Kabbani, Ossama; Tajima, Kazuo; Ikari, Akira; Hara, Akira

    2014-01-01

    Inhalation of 9,10-phenanthrenequinone (9,10-PQ), a major quinone in diesel exhaust, exerts fatal damage against a variety of cells involved in respiratory function. Here, we show that treatment with high concentrations of 9,10-PQ evokes apoptosis of lung cancer A549 cells through production of reactive oxygen species (ROS). In contrast, 9,10-PQ at its concentrations of 2 and 5 μM elevated the potentials for proliferation, invasion, metastasis and tumorigenesis, all of which were almost completely inhibited by addition of an antioxidant N-acetyl-L-cysteine, inferring a crucial role of ROS in the overgrowth and malignant progression of lung cancer cells. Comparison of mRNA expression levels of six aldo-keto reductases (AKRs) in the 9,10-PQ-treated cells advocated up-regulation of AKR1B10 as a major cause contributing to the lung cancer malignancy. In support of this, the elevation of invasive, metastatic and tumorigenic activities in the 9,10-PQ-treated cells was significantly abolished by the addition of a selective AKR1B10 inhibitor oleanolic acid. Intriguingly, zymographic and real-time PCR analyses revealed remarkable increases in secretion and expression, respectively, of matrix metalloproteinase 2 during the 9,10-PQ treatment, and suggested that the AKR1B10 up-regulation and resultant activation of mitogen-activated protein kinase cascade are predominant mechanisms underlying the metalloproteinase induction. In addition, HPLC analysis and cytochrome c reduction assay in in vitro 9,10-PQ reduction by AKR1B10 demonstrated that the enzyme catalyzes redox-cycling of this quinone, by which ROS are produced. Collectively, these results suggest that AKR1B10 is a key regulator involved in overgrowth and malignant progression of the lung cancer cells through ROS production due to 9,10-PQ redox-cycling. - Highlights: • 9,10-PQ promotes invasion, metastasis and tumorigenicity in lung cancer cells. • The 9,10-PQ-elicited promotion is possibly due to AKR1B10 up

  5. Exposure to 9,10-phenanthrenequinone accelerates malignant progression of lung cancer cells through up-regulation of aldo-keto reductase 1B10

    Energy Technology Data Exchange (ETDEWEB)

    Matsunaga, Toshiyuki, E-mail: matsunagat@gifu-pu.ac.jp [Laboratory of Biochemistry, Gifu Pharmaceutical University, Gifu 501-1196 (Japan); Morikawa, Yoshifumi; Haga, Mariko; Endo, Satoshi [Laboratory of Biochemistry, Gifu Pharmaceutical University, Gifu 501-1196 (Japan); Soda, Midori; Yamamura, Keiko [Laboratory of Clinical Pharmacy, School of Pharmacy, Aichi Gakuin University, Nagoya 464-8650 (Japan); El-Kabbani, Ossama [Monash Institute of Pharmaceutical Sciences, Monash University, Victoria 3052 (Australia); Tajima, Kazuo [Faculty of Pharmaceutical Sciences, Hokuriku University, Kanazawa 920-1181 (Japan); Ikari, Akira [Laboratory of Biochemistry, Gifu Pharmaceutical University, Gifu 501-1196 (Japan); Hara, Akira [Faculty of Engineering, Gifu University, Gifu 501-1193 (Japan)

    2014-07-15

    Inhalation of 9,10-phenanthrenequinone (9,10-PQ), a major quinone in diesel exhaust, exerts fatal damage against a variety of cells involved in respiratory function. Here, we show that treatment with high concentrations of 9,10-PQ evokes apoptosis of lung cancer A549 cells through production of reactive oxygen species (ROS). In contrast, 9,10-PQ at its concentrations of 2 and 5 μM elevated the potentials for proliferation, invasion, metastasis and tumorigenesis, all of which were almost completely inhibited by addition of an antioxidant N-acetyl-L-cysteine, inferring a crucial role of ROS in the overgrowth and malignant progression of lung cancer cells. Comparison of mRNA expression levels of six aldo-keto reductases (AKRs) in the 9,10-PQ-treated cells advocated up-regulation of AKR1B10 as a major cause contributing to the lung cancer malignancy. In support of this, the elevation of invasive, metastatic and tumorigenic activities in the 9,10-PQ-treated cells was significantly abolished by the addition of a selective AKR1B10 inhibitor oleanolic acid. Intriguingly, zymographic and real-time PCR analyses revealed remarkable increases in secretion and expression, respectively, of matrix metalloproteinase 2 during the 9,10-PQ treatment, and suggested that the AKR1B10 up-regulation and resultant activation of mitogen-activated protein kinase cascade are predominant mechanisms underlying the metalloproteinase induction. In addition, HPLC analysis and cytochrome c reduction assay in in vitro 9,10-PQ reduction by AKR1B10 demonstrated that the enzyme catalyzes redox-cycling of this quinone, by which ROS are produced. Collectively, these results suggest that AKR1B10 is a key regulator involved in overgrowth and malignant progression of the lung cancer cells through ROS production due to 9,10-PQ redox-cycling. - Highlights: • 9,10-PQ promotes invasion, metastasis and tumorigenicity in lung cancer cells. • The 9,10-PQ-elicited promotion is possibly due to AKR1B10 up

  6. Ordinary mode auroral kilometric radiation fine structure observed by DE 1

    International Nuclear Information System (INIS)

    Benson, R.F.; Mellott, M.M.; Huff, R.L.; Gurnett, D.A.

    1988-01-01

    The fine structure observed with intense right-hand extraordinary (R-X) mode auroral kilometric radiation (AKR) has received major theoretical attention. Data from the Dynamics Explorer 1 plasma wave instrument indicate that left-hand ordinary (L-O) mode AKR posses similar fine structure. Several theories have been proposed to explain the fine structure of the R-X mode AKR. In order to account for the L-O mode fine structure, these theories will have to be modified to produce the L-O mode directly or will have to rely on mode conversion processes from the R-X to the L-O mode

  7. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  8. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  9. Aldo-keto reductase 1B10 promotes development of cisplatin resistance in gastrointestinal cancer cells through down-regulating peroxisome proliferator-activated receptor-γ-dependent mechanism.

    Science.gov (United States)

    Matsunaga, Toshiyuki; Suzuki, Ayaka; Kezuka, Chihiro; Okumura, Naoko; Iguchi, Kazuhiro; Inoue, Ikuo; Soda, Midori; Endo, Satoshi; El-Kabbani, Ossama; Hara, Akira; Ikari, Akira

    2016-08-25

    Cisplatin (cis-diamminedichloroplatinum, CDDP) is one of the most effective chemotherapeutic drugs that are used for treatment of patients with gastrointestinal cancer cells, but its continuous administration often evokes the development of chemoresistance. In this study, we investigated alterations in antioxidant molecules and functions using a newly established CDDP-resistant variant of gastric cancer MKN45 cells, and found that aldo-keto reductase 1B10 (AKR1B10) is significantly up-regulated with acquisition of the CDDP resistance. In the nonresistant MKN45 cells, the sensitivity to cytotoxic effect of CDDP was decreased and increased by overexpression and silencing of AKR1B10, respectively. In addition, the AKR1B10 overexpression markedly suppressed accumulation and cytotoxicity of 4-hydroxy-2-nonenal that is produced during lipid peroxidation by CDDP treatment, suggesting that the enzyme acts as a crucial factor for facilitation of the CDDP resistance through inhibiting induction of oxidative stress by the drug. Transient exposure to CDDP and induction of the CDDP resistance decreased expression of peroxisome proliferator-activated receptor-γ (PPARγ) in MKN45 and colon cancer LoVo cells. Additionally, overexpression of PPARγ in the cells elevated the sensitivity to the CDDP toxicity, which was further augmented by concomitant treatment with a PPARγ ligand rosiglitazone. Intriguingly, overexpression of AKR1B10 in the cells resulted in a decrease in PPARγ expression, which was recovered by addition of an AKR1B10 inhibitor oleanolic acid, inferring that PPARγ is a downstream target of AKR1B10-dependent mechanism underlying the CDDP resistance. Combined treatment with the AKR1B10 inhibitor and PPARγ ligand elevated the CDDP sensitivity, which was almost the same level as that in the parental cells. These results suggest that combined treatment with the AKR1B10 inhibitor and PPARγ ligand is an effective adjuvant therapy for overcoming CDDP resistance of

  10. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  11. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  12. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  13. Auroral kilometric radiation source region observations from ISIS 1

    International Nuclear Information System (INIS)

    Benson, R.F.

    1981-01-01

    The ISIS 1 observations of the high-frequency portion of the auroral kilometric radiation (AKR) spectrum are considered, that is, from the minimum frequency encountered for the extraordinary mode cut-off (approximately 450 kHz) to the upper frequency cut-off (approximately 800 kHz). AKR is found to be generated in the extraordinary mode just above the local cutoff frequency and to emanate in a direction that is nearly perpendicular to the magnetic field. It occurs within local depletions of electron density, where the ratio of plasma frequency to cyclotron frequency is below 0.2. The density depletion is restricted to altitudes above approximately 2,000 km, and the upper AKR frequency limit corresponds to the extraordinary cutoff frequency at this altitude

  14. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  15. Purification, cloning, functional expression and characterization of perakine reductase: the first example from the AKR enzyme family, extending the alkaloidal network of the plant Rauvolfia.

    Science.gov (United States)

    Sun, Lianli; Ruppert, Martin; Sheludko, Yuri; Warzecha, Heribert; Zhao, Yu; Stöckigt, Joachim

    2008-07-01

    Perakine reductase (PR) catalyzes an NADPH-dependent step in a side-branch of the 10-step biosynthetic pathway of the alkaloid ajmaline. The enzyme was cloned by a "reverse-genetic" approach from cell suspension cultures of the plant Rauvolfia serpentina (Apocynaceae) and functionally expressed in Escherichia coli as the N-terminal His(6)-tagged protein. PR displays a broad substrate acceptance, converting 16 out of 28 tested compounds with reducible carbonyl function which belong to three substrate groups: benzaldehyde, cinnamic aldehyde derivatives and monoterpenoid indole alkaloids. The enzyme has an extraordinary selectivity in the group of alkaloids. Sequence alignments define PR as a new member of the aldo-keto reductase (AKR) super family, exhibiting the conserved catalytic tetrad Asp52, Tyr57, Lys84, His126. Site-directed mutagenesis of each of these functional residues to an alanine residue results in >97.8% loss of enzyme activity, in compounds of each substrate group. PR represents the first example of the large AKR-family which is involved in the biosynthesis of plant monoterpenoid indole alkaloids. In addition to a new esterase, PR significantly extends the Rauvolfia alkaloid network to the novel group of peraksine alkaloids.

  16. Dysregulated Intrahepatic CD4+ T-Cell Activation Drives Liver Inflammation in Ileitis-Prone SAMP1/YitFc MiceSummary

    Directory of Open Access Journals (Sweden)

    Sara Omenetti

    2015-07-01

    Full Text Available Background & Aims: Liver inflammation is a common extraintestinal manifestation of inflammatory bowel disease (IBD, but whether liver involvement is a consequence of a primary intestinal defect or results from alternative pathogenic processes remains unclear. Therefore, we sought to determine the potential pathogenic mechanism(s of concomitant liver inflammation in an established murine model of IBD. Methods: Liver inflammation and immune cell subsets were characterized in ileitis-prone SAMP1/YitFc (SAMP and AKR/J (AKR control mice, lymphocyte-depleted SAMP (SAMPxRag-1−/−, and immunodeficient SCID recipient mice receiving SAMP or AKR donor CD4+ T cells. Proliferation and suppressive capacity of CD4+ T-effector (Teff and T-regulatory (Treg cells from gut-associated lymphoid tissue (GALT and livers of SAMP and AKR mice were measured. Results: Surprisingly, prominent inflammation was detected in 4-week-old SAMP livers before histologic evidence of ileitis, whereas both disease phenotypes were absent in age-matched AKR mice. SAMP liver disease was characterized by abundant infiltration of lymphocytes, required for hepatic inflammation to occur, a TH1-skewed environment, and phenotypically activated CD4+ T cells. SAMP intrahepatic CD4+ T cells also had the ability to induce liver and ileal inflammation when adoptively transferred into SCID recipients, whereas GALT-derived CD4+ T cells produced milder ileitis but not liver inflammation. Interestingly, SAMP intrahepatic CD4+ Teff cells showed increased proliferation compared with both SAMP GALT- and AKR liver-derived CD4+ Teff cells, and SAMP intrahepatic Tregs were decreased among CD4+ T cells and impaired in in vitro suppressive function compared with AKR. Conclusions: Activated intrahepatic CD4+ T cells induce liver inflammation and contribute to experimental ileitis via locally impaired hepatic immunosuppressive function. Keywords: Hepatic CD4+ T Cells, IBD-Associated Liver

  17. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  18. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  19. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  20. Occupational analysis for the Angra-1 reactor

    International Nuclear Information System (INIS)

    Moraes, A.

    1991-01-01

    Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

  1. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  2. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  3. Aldo-keto reductase enzymes detoxify glyphosate and improve herbicide resistance in plants.

    Science.gov (United States)

    Vemanna, Ramu S; Vennapusa, Amaranatha Reddy; Easwaran, Murugesh; Chandrashekar, Babitha K; Rao, Hanumantha; Ghanti, Kirankumar; Sudhakar, Chinta; Mysore, Kirankumar S; Makarla, Udayakumar

    2017-07-01

    In recent years, concerns about the use of glyphosate-resistant crops have increased because of glyphosate residual levels in plants and development of herbicide-resistant weeds. In spite of identifying glyphosate-detoxifying genes from microorganisms, the plant mechanism to detoxify glyphosate has not been studied. We characterized an aldo-keto reductase gene from Pseudomonas (PsAKR1) and rice (OsAKR1) and showed, by docking studies, both PsAKR1 and OsAKR1 can efficiently bind to glyphosate. Silencing AKR1 homologues in rice and Nicotiana benthamiana or mutation of AKR1 in yeast and Arabidopsis showed increased sensitivity to glyphosate. External application of AKR proteins rescued glyphosate-mediated cucumber seedling growth inhibition. Regeneration of tobacco transgenic lines expressing PsAKR1 or OsAKRI on glyphosate suggests that AKR can be used as selectable marker to develop transgenic crops. PsAKR1- or OsAKRI-expressing tobacco and rice transgenic plants showed improved tolerance to glyphosate with reduced accumulation of shikimic acid without affecting the normal photosynthetic rates. These results suggested that AKR1 when overexpressed detoxifies glyphosate in planta. © 2016 The Authors. Plant Biotechnology Journal published by Society for Experimental Biology and The Association of Applied Biologists and John Wiley & Sons Ltd.

  4. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  5. Bone marrow transplantation immunology

    International Nuclear Information System (INIS)

    Trentin, J.J.; Kiessling, R.; Wigzell, H.; Gallagher, M.T.; Datta, S.K.; Kulkarni, S.S.

    1977-01-01

    Tests were made to determine whether genetic resistance (GR) to bone marrow transplantation represents a natural lymphoma-leukemia defense mechanism, as follows: (C57 x AKR) F 1 hybrid mice show GR to C57 parental bone marrow cells, but not to AKR parental bone marrow cells (C3H x AKR) F 1 hybrids show no GR to bone marrow transplantation from either parental strain. However, transplantation of AKR lymphoma cells into lethally irradiated ''resistant'' (C57 x AKR) F 1 and ''nonresistant'' (C3H x AKR) F 1 hybrids produced lymphomatous spleen colonies in ''nonresistant'' hybrids but not in ''resistant'' hybrids. Thus ''resistant'' (C57 x AKR) F 1 hybrids can recognize and reject AKR lymphoma cells, but not normal AKR bone marrow cells. A normal biologic role of leukemia-lymphoma surveillance was postulated for genetic resistance to marrow transplantation, directed at antigens which, like TL, are expressed on normal hemopoietic cells of some strains, but only on leukemic cells of other strains

  6. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  7. The aldo-keto reductase superfamily homepage.

    Science.gov (United States)

    Hyndman, David; Bauman, David R; Heredia, Vladi V; Penning, Trevor M

    2003-02-01

    The aldo-keto reductases (AKRs) are one of the three enzyme superfamilies that perform oxidoreduction on a wide variety of natural and foreign substrates. A systematic nomenclature for the AKR superfamily was adopted in 1996 and was updated in September 2000 (visit www.med.upenn.edu/akr). Investigators have been diligent in submitting sequences of functional proteins to the Web site. With the new additions, the superfamily contains 114 proteins expressed in prokaryotes and eukaryotes that are distributed over 14 families (AKR1-AKR14). The AKR1 family contains the aldose reductases, the aldehyde reductases, the hydroxysteroid dehydrogenases and steroid 5beta-reductases, and is the largest. Other families of interest include AKR6, which includes potassium channel beta-subunits, and AKR7 the aflatoxin aldehyde reductases. Two new families include AKR13 (yeast aldose reductase) and AKR14 (Escherichia coli aldehyde reductase). Crystal structures of many AKRs and their complexes with ligands are available in the PDB and accessible through the Web site. Each structure has the characteristic (alpha/beta)(8)-barrel motif of the superfamily, a conserved cofactor binding site and a catalytic tetrad, and variable loop structures that define substrate specificity. Although the majority of AKRs are monomeric proteins of about 320 amino acids in length, the AKR2, AKR6 and AKR7 family may form multimers. To expand the nomenclature to accommodate multimers, we recommend that the composition and stoichiometry be listed. For example, AKR7A1:AKR7A4 (1:3) would designate a tetramer of the composition indicated. The current nomenclature is recognized by the Human Genome Project (HUGO) and the Web site provides a link to genomic information including chromosomal localization, gene boundaries, human ESTs and SNPs and much more.

  8. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  9. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  10. Targeting the Mevalonate Pathway to Reduce Mortality from Ovarian Cancer

    Science.gov (United States)

    2015-10-01

    RHOB, RRAS, RHOA, PLAU, RHOF, ITGB5, ITGB3 Methylglyoxal degradation III 2.17Eþ00 1.30E01 AKR1C1/AKR1C2, AKR1C3, AKR1C4 Dopamine degradation...2007;25: 2921–7. 6. Bober SL, Recklitis CJ, Bakan J, Garber JE, Patenaude AF. Addressing sexual dysfunction after risk-reducing salpingo-oophorectomy

  11. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  12. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  13. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  14. BIOLOGICAL ROLE OF ALDO-KETO REDUCTASES IN RETINOIC ACID BIOSYNTHESIS AND SIGNALING

    Directory of Open Access Journals (Sweden)

    F. Xavier eRuiz

    2012-04-01

    Full Text Available Several aldo-keto reductase (AKR enzymes from subfamilies 1B and 1C show retinaldehyde reductase activity, having low Km and kcat values. Only AKR1B10 and 1B12, with all-trans-retinaldehyde, and AKR1C3, with 9-cis-retinaldehyde, display high catalytic efficiency. Major structural determinants for retinaldehyde isomer specificity are located in the external loops (A and C for AKR1B10, and B for AKR1C3, as assessed by site-directed mutagenesis and molecular dynamics. Cellular models have shown that AKR1B and 1C enzymes are well suited to work in vivo as retinaldehyde reductases and to regulate retinoic acid (RA biosynthesis at hormone pre-receptor level. An additional physiological role for the retinaldehyde reductase activity of these enzymes, consistent with their tissue localization, is their participation in β-carotene absorption. Retinaldehyde metabolism may be subjected to subcellular compartmentalization, based on enzyme localization. While retinaldehyde oxidation to RA takes place in the cytosol, reduction to retinol could take place in the cytosol by AKRs or in the membranes of endoplasmic reticulum by microsomal retinaldehyde reductases. Upregulation of some AKR1 enzymes in different cancer types may be linked to their induction by oxidative stress and to their participation in different signaling pathways related to cell proliferation. AKR1B10 and AKR1C3, through their retinaldehyde reductase activity, trigger a decrease in the RA biosynthesis flow, resulting in RA deprivation and consequently lower differentiation, with an increased cancer risk in target tissues. Rational design of selective AKR inhibitors could lead to development of novel drugs for cancer treatment as well as reduction of chemotherapeutic drug resistance.

  15. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  16. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  17. Modernization and Refurbishment of the RECH-1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

    2014-08-15

    The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

  18. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    Rataj, J.

    2010-01-01

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  19. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  20. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  1. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

    1994-01-01

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  2. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  3. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    Gonzalez, J. M.

    1993-01-01

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  4. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  5. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    Haskin, F.E.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  6. Perspectives on reactor safety. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  7. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  8. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  9. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    Perry, E.F.

    1995-01-01

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

  10. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  11. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  12. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

  13. New instrumentation for the IPR-R1 reactor of CDTN

    International Nuclear Information System (INIS)

    Carvalho, P.V.R. de.

    1992-01-01

    The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

  14. Preliminary Evaluation on Characteristics of Waste from Pyro-processing (FS v5.1)

    International Nuclear Information System (INIS)

    Kim, In-Young; Choi, Heui-Joo

    2016-01-01

    In this study, characteristic of waste from pyro-processing which based on material balance FS v5.1 is evaluated for revising of A-KRS concept. To reduce volume and toxicity of PWR SNFs, the P and T Technology using pyro-processing and SFR is under development in KAERI. In accordance with this R and D, the A-KRS was developed by KAERI for disposal of waste from pyro-processing. After A-KRS concept development, material balance has been revised and characteristics of waste have been changed. To solve impending saturation of storage capacity of NPP sites, national policy on SNF management seems to be determined shortly. Demand on detailed analysis on impact of P and T using pyro-processing and SFR as base data is increasing. To compare direct disposal scenario to the P and T scenario using pyro-processing and SFR, updating of A-KRS reflecting amendment of material balance is required. In this study, characteristic of waste from pyro-processing which based on material balance FS v5.1 is evaluated. Every 15 types of outputs generated from pyro-processing are evaluated. Great reduction can be achieved by complete reuse of U/TRU/RE in SFR, because most of decay heat, radioactivity and radiotoxicity are generated from U/TRU/RE ingot. Within about 300 years, the fly ash filter containing Cs, Sr containing salt waste are important waste in perspective of decay heat and radioactivity. Importance of RE containing salt waste reduced. Management of hold-ups must be clarified because hold-ups take large portion of decay heat, radioactivity in the material balance FS v5.1

  15. Preliminary Evaluation on Characteristics of Waste from Pyro-processing (FS v5.1)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In-Young; Choi, Heui-Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, characteristic of waste from pyro-processing which based on material balance FS v5.1 is evaluated for revising of A-KRS concept. To reduce volume and toxicity of PWR SNFs, the P and T Technology using pyro-processing and SFR is under development in KAERI. In accordance with this R and D, the A-KRS was developed by KAERI for disposal of waste from pyro-processing. After A-KRS concept development, material balance has been revised and characteristics of waste have been changed. To solve impending saturation of storage capacity of NPP sites, national policy on SNF management seems to be determined shortly. Demand on detailed analysis on impact of P and T using pyro-processing and SFR as base data is increasing. To compare direct disposal scenario to the P and T scenario using pyro-processing and SFR, updating of A-KRS reflecting amendment of material balance is required. In this study, characteristic of waste from pyro-processing which based on material balance FS v5.1 is evaluated. Every 15 types of outputs generated from pyro-processing are evaluated. Great reduction can be achieved by complete reuse of U/TRU/RE in SFR, because most of decay heat, radioactivity and radiotoxicity are generated from U/TRU/RE ingot. Within about 300 years, the fly ash filter containing Cs, Sr containing salt waste are important waste in perspective of decay heat and radioactivity. Importance of RE containing salt waste reduced. Management of hold-ups must be clarified because hold-ups take large portion of decay heat, radioactivity in the material balance FS v5.1.

  16. Crystal structure of conjugated polyketone reductase (CPR-C1) from Candida parapsilosis IFO 0708 complexed with NADPH.

    Science.gov (United States)

    Qin, Hui-Min; Yamamura, Akihiro; Miyakawa, Takuya; Kataoka, Michihiko; Maruoka, Shintaro; Ohtsuka, Jun; Nagata, Koji; Shimizu, Sakayu; Tanokura, Masaru

    2013-11-01

    Conjugated polyketone reductase (CPR-C1) from Candida parapsilosis IFO 0708 is a member of the aldo-keto reductase (AKR) superfamily and reduces ketopantoyl lactone to d-pantoyl lactone in a NADPH-dependent and stereospecific manner. We determined the crystal structure of CPR-C1.NADPH complex at 2.20 Å resolution. CPR-C1 adopted a triose-phosphate isomerase (TIM) barrel fold at the core of the structure in which Thr25 and Lys26 of the GXGTX motif bind uniquely to the adenosine 2'-phosphate group of NADPH. This finding provides a novel structural basis for NADPH binding of the AKR superfamily. Copyright © 2013 Wiley Periodicals, Inc.

  17. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  18. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

    2002-01-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  19. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  20. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    1988-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  1. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    1978-04-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  2. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  3. Assessing Options for Contingent Contracting of Merchant Ships for Naval and Expeditionary Operations

    Science.gov (United States)

    2008-12-01

    be bales, boxes, machinery, 200 tones of steel products, 1,000 tones of soya beans and 2,000 pallets of bottled water. However, we could categorize...vehicles and helicopters. When not activated, the ships are layberthed in the following locations [21]: • USNS Algol (T-AKR 287) Violet, La . • USNS...Bellatrix (T-AKR 288) Marrero, La . • USNS Denebola (T-AKR 289) Norfolk, Va. • USNS Pollux (T-AKR 290) Violet, La . • USNS Altair (T-AKR 291

  4. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  5. Effects of resveratrol on rat neurosteroid synthetic enzymes.

    Science.gov (United States)

    Wang, Yiluan; Sun, Jianliang; Chen, Ling; Zhou, Songyi; Lin, Han; Wang, Yiyan; Lin, Nengming; Ge, Ren-Shan

    2017-10-01

    Resveratrol, a common polyphenol, has extensive pharmacological activities. Resveratrol inhibits some steroid biosynthetic enzymes, indicating that it may block neurosteroid synthesis. The objective of the present study is to investigate the inhibition of resveratrol on neurosteroidogenic enzymes rat 5α-reductase 1 (SRD5A1), 3α-hydroxysteroid dehydrogenase (AKR1C9), and retinol dehydrogenase 2 (RDH2). The IC 50 values of resveratrol on SRD5A1, AKR1C9, and RDH2 were >100μM, 0.436±0.070μM, and 4.889±0.062μM, respectively. Resveratrol competitively inhibited rat AKR1C9 and RDH2 against steroid substrates. Docking showed that resveratrol bound to the steroid binding pocket of AKR1C9. It exerted a mixed mode on these AKR1C9 and RDH2 against cofactors. In conclusion, resveratrol potently inhibited rat AKR1C9 and RDH2 to regulate local neurosteroid levels. Copyright © 2017. Published by Elsevier B.V.

  6. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  7. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  8. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  9. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  10. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  11. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  12. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  13. Department of Reactor Technology: annual progress report 1 January - 31 December 1976

    International Nuclear Information System (INIS)

    1977-06-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

  14. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  15. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  16. Department of Reactor Technology annual progress report 1 January - 31 December 1978

    International Nuclear Information System (INIS)

    1979-04-01

    The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  17. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  18. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    1994-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  19. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  20. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  1. Modification of the IAN-R1 reactor

    International Nuclear Information System (INIS)

    Jaime, J.; Ahumada, S.; Spin, R.A.

    1990-01-01

    The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

  2. Sex hormones reduce NNK detoxification through inhibition of short-chain dehydrogenases/reductases and aldo-keto reductases in vitro.

    Science.gov (United States)

    Stapelfeld, Claudia; Maser, Edmund

    2017-10-01

    Carbonyl reduction is an important metabolic pathway for endogenous and xenobiotic substances. The tobacco specific nitrosamine 4-(methylnitrosamino)-1-(3-pyridyl)-1-butanone (NNK, nicotine-derived nitrosamine ketone) is classified as carcinogenic to humans (IARC, Group 1) and considered to play the most important role in tobacco-related lung carcinogenesis. Detoxification of NNK through carbonyl reduction is catalyzed by members of the AKR- and the SDR-superfamilies which include AKR1B10, AKR1C1, AKR1C2, AKR1C4, 11β-HSD1 and CBR1. Because some reductases are also involved in steroid metabolism, five different hormones were tested for their inhibitory effect on NNK carbonyl reduction. Two of those hormones were estrogens (estradiol and ethinylestradiol), another two hormones belong to the gestagen group (progesterone and drospirenone) and the last tested hormone was an androgen (testosterone). Furthermore, one of the estrogens (ethinylestradiol) and one of the gestagens (drospirenone) are synthetic hormones, used as hormonal contraceptives. Five of six NNK reducing enzymes (AKR1B10, AKR1C1, AKR1C2, AKR1C4 and 11β-HSD1) were significantly inhibited by the tested sex hormones. Only NNK reduction catalyzed by CBR1 was not significantly impaired. In the case of the other five reductases, gestagens had remarkably stronger inhibitory effects at a concentration of 25 μM (progesterone: 66-88% inhibition; drospirenone: 26-87% inhibition) in comparison to estrogens (estradiol: 17-51% inhibition; ethinylestradiol: 14-79% inhibition) and androgens (14-78% inhibition). Moreover, in most cases the synthetic hormones showed a greater ability to inhibit NNK reduction than the physiologic derivatives. These results demonstrate that male and female sex hormones have different inhibitory potentials, thus indicating that there is a varying detoxification capacity of NNK in men and women which could result in a different risk for developing lung cancer. Copyright © 2017 Elsevier B

  3. Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Kassar, E.

    1986-01-01

    The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

  4. Mapping of auroral kilometric radiation sources to the aurora

    International Nuclear Information System (INIS)

    Huff, R.L.; Calvert, W.; Craven, J.D.; Frank, L.A.; Gurnett, D.A.

    1988-01-01

    Auroral kilometric radiation (AKR) and optical auroral emissions are observed simultaneously using plasma wave instrumentation and auroral imaging photometers acrried on the DE 1 spacecraft. The DE 1 plasma wave instrument measures the relative phase of signals from orthogonal electric dipole antennas, and from these measurements, apparent source directions can be determined with a high degree of precision. Wave data are analyzed for several strong AKR events, and source directions are determined for several emission frequencies. By assuming that the AKR originates at cyclotron resonant altitudes, a condidate source field line is identified. When the selected source field line is traced down to auroral altitudes on the concurrent DE 1 auroral image, a striking correspondence between the AKR source field line and localized auroral features is produced. The magnetic mapping study provides strong evidence that AKR sources occur on field lines associated with discrete auroral arcs, and it provides confirmation that AKR is generated near the electron cyclotron frequency

  5. Activation of pregnane X receptor by pregnenolone 16 α-carbonitrile prevents high-fat diet-induced obesity in AKR/J mice.

    Directory of Open Access Journals (Sweden)

    Yongjie Ma

    Full Text Available Pregnane X receptor (PXR is known to function as a xenobiotic sensor to regulate xenobiotic metabolism through selective transcription of genes responsible for maintaining physiological homeostasis. Here we report that the activation of PXR by pregnenolone 16α-carbonitrile (PCN in AKR/J mice can prevent the development of high-fat diet-induced obesity and insulin resistance. The beneficial effects of PCN treatment are seen with reduced lipogenesis and gluconeogenesis in the liver, and lack of hepatic accumulation of lipid and lipid storage in the adipose tissues. RT-PCR analysis of genes involved in gluconeogenesis, lipid metabolism and energy homeostasis reveal that PCN treatment on high-fat diet-fed mice reduces expression in the liver of G6Pase, Pepck, Cyp7a1, Cd36, L-Fabp, Srebp, and Fas genes and slightly enhances expression of Cyp27a1 and Abca1 genes. RT-PCR analysis of genes involved in adipocyte differentiation and lipid metabolism in white adipose tissue show that PCN treatment reduces expression of Pparγ2, Acc1, Cd36, but increases expression of Cpt1b and Pparα genes in mice fed with high-fat diet. Similarly, PCN treatment of animals on high-fat diet increases expression in brown adipose tissue of Pparα, Hsl, Cpt1b, and Cd36 genes, but reduces expression of Acc1 and Scd-1 genes. PXR activation by PCN in high-fat diet fed mice also increases expression of genes involved in thermogenesis in brown adipose tissue including Dio2, Pgc-1α, Pgc-1β, Cidea, and Ucp-3. These results verify the important function of PXR in lipid and energy metabolism and suggest that PXR represents a novel therapeutic target for prevention and treatment of obesity and insulin resistance.

  6. Expression of mink cell focus-forming murine leukemia virus-related transcripts in AKR mice

    International Nuclear Information System (INIS)

    Khan, A.S.; Laigret, F.; Rodi, C.P.

    1987-01-01

    The authors used a synthetic 16-base-pair mink cell focus-forming (MCF) env-specific oligomer as radiolabeled probe to study MCF murine leukemia virus (MuLV)-related transcripts in brain, kidney, liver, spleen, and thymus tissues of AKR mice ranging from 5 weeks to 6 months (mo) of age. Tissue-specific expression of poly(A) + RNAs was seen. In addition, all the tissues tested contained 3.0-kb messages. The transcription of these MCF-related mRNAs was independent of the presence of ecotropic and xenotropic MuLVs. In general, expression of the MCF env-related transcripts appeared to peak at 2 mo of age; these messages were barely detectable in brain, kidney, liver, and spleen tissues after 2 mo and in thymus tissue after 4 mo of age. All of the subgenomic MCF env-related mRNAs appeared to contain the 190-base-pair cellular DNA insert, characteristic of the long terminal repeats associated with endogenous MCF env-related proviruses. No genomic-size (8.4-kb) transcripts corresponding to endogenous MCF-related proviruses were detected. An 8.4-kb MCF env-related mRNA was first seen at 3 mo of age, exclusively in thymus tissue. This species most likely represents the first appearance of a recombinant MCF-related MuLV genome. The transcripts which were detected in thymus tissue might be involved in the generation of leukemogenic MCF viruses

  7. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  8. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  9. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  10. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  11. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  12. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  13. Spontaneous autoimmune gastritis and hypochlorhydria are manifest in the ileitis-prone SAMP1/YitFcs mice.

    Science.gov (United States)

    Ernst, P B; Erickson, L D; Loo, W M; Scott, K G; Wiznerowicz, E B; Brown, C C; Torres-Velez, F J; Alam, M S; Black, S G; McDuffie, M; Feldman, S H; Wallace, J L; McKnight, G W; Padol, I T; Hunt, R H; Tung, K S

    2012-01-01

    SAMP1/YitFcs mice serve as a model of Crohn's disease, and we have used them to assess gastritis. Gastritis was compared in SAMP1/YitFcs, AKR, and C57BL/6 mice by histology, immunohistochemistry, and flow cytometry. Gastric acid secretion was measured in ligated stomachs, while anti-parietal cell antibodies were assayed by immunofluorescence and enzyme-linked immunosorbent spot assay. SAMP1/YitFcs mice display a corpus-dominant, chronic gastritis with multifocal aggregates of mononuclear cells consisting of T and B lymphocytes. Relatively few aggregates were observed elsewhere in the stomach. The infiltrates in the oxyntic mucosa were associated with the loss of parietal cell mass. AKR mice, the founder strain of the SAMP1/YitFcs, also have gastritis, although they do not develop ileitis. Genetic studies using SAMP1/YitFcs-C57BL/6 congenic mice showed that the genetic regions regulating ileitis had comparable effects on gastritis. The majority of the cells in the aggregates expressed the T cell marker CD3 or the B cell marker B220. Adoptive transfer of SAMP1/YitFcs CD4(+) T helper cells, with or without B cells, into immunodeficient recipients induced a pangastritis and duodenitis. SAMP1/YitFcs and AKR mice manifest hypochlorhydria and anti-parietal cell antibodies. These data suggest that common genetic factors controlling gastroenteric disease in SAMP1/YitFcs mice regulate distinct pathogenic mechanisms causing inflammation in separate sites within the digestive tract.

  14. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    Festinesi, A.

    1986-01-01

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  15. Neutronic studies in the enrichment reduction of research reactor IEAR-1

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

    1987-01-01

    In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

  16. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    1980-01-01

    prior to the beginning of the course was of particular value. Interesting scientific visits and demonstrations at the Isotope Institute and at the Central Research Institute for Physics (IFKI), both of the Hungarian Academy of Sciences, were also arranged. During the Study Tour at the Central Institute for Nuclear Research in Rossendorf near Dresden, German Democratic Republic, the participants had the opportunity to observe the organization of a 10 MW nuclear reactor where radioisotopes and radiopharmaceuticals are produced on a commercial scale. Lectures were delivered by local scientists on some of their programmes in applied research in solid state physics and material sciences. At the Technical University of Dresden, the group visited the homogeneous solid-moderated zero-power training reactor (AKR), primarily dedicated to nuclear education and training. Studies on different theoretical and experimental aspects of radiation protection (solid state nuclear track and thermoluminescent detectors) are also being carried out. The last day of the Study Tour was devoted to a visit to the College for Advanced Technology at Zittau, where a training reactor with a power of a few watts has been recently installed. (author)

  17. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  18. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  19. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  20. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  1. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  2. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  3. Ageing problems and renovation programme of ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Khattab, M.S.; Sultan, M.A.

    1995-01-01

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  4. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  5. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  6. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  7. Occurrence of spontaneous periodontal disease in the SAMP1/YitFc murine model of Crohn disease.

    Science.gov (United States)

    Pietropaoli, Davide; Del Pinto, Rita; Corridoni, Daniele; Rodriguez-Palacios, Alexander; Di Stefano, Gabriella; Monaco, Annalisa; Weinberg, Aaron; Cominelli, Fabio

    2014-12-01

    Oral involvement is often associated with inflammatory bowel disease (IBD). Recent evidence suggests a high incidence of periodontal disease in patients with Crohn disease (CD). To the best of the authors' knowledge, no animal model of IBD that displays associated periodontal disease was reported previously. The aim of this study is to investigate the occurrence and progression of periodontal disease in SAMP1/YitFc (SAMP) mice that spontaneously develop a CD-like ileitis. In addition, the temporal correlation between the onset and progression of periodontal disease and the onset of ileitis in SAMP mice was studied. At different time points, SAMP and parental AKR/J (AKR) control mice were sacrificed, and mandibles were prepared for stereomicroscopy and histology. Terminal ilea were collected for histologic assessment of inflammation score. Periodontal status, i.e., alveolar bone loss (ABL) and alveolar bone crest, was examined by stereomicroscopy and histomorphometry, respectively. ABL increased in both strains with age. SAMP mice showed greater ABL compared with AKR mice by 12 weeks of age, with maximal differences observed at 27 weeks of age. AKR control mice did not show the same severity of periodontal disease. Interestingly, a strong positive correlation was found between ileitis severity and ABL in SAMP mice, independent of age. The present results demonstrate the occurrence of periodontal disease in a mouse model of progressive CD-like ileitis. In addition, the severity of periodontitis strongly correlated with the severity of ileitis, independent of age, suggesting that common pathogenic mechanisms, such as abnormal immune response and dysbiosis, may be shared between these two phenotypes.

  8. Spontaneous autoimmune gastritis and hypochlorhydria are manifest in the ileitis-prone SAMP1/YitFcs mice

    Science.gov (United States)

    Erickson, L. D.; Loo, W. M.; Scott, K. G.; Wiznerowicz, E. B.; Brown, C. C.; Torres-Velez, F. J.; Alam, M. S.; Black, S. G.; McDuffie, M.; Feldman, S. H.; Wallace, J. L.; McKnight, G. W.; Padol, I. T.; Hunt, R. H.; Tung, K. S.

    2012-01-01

    SAMP1/YitFcs mice serve as a model of Crohn's disease, and we have used them to assess gastritis. Gastritis was compared in SAMP1/YitFcs, AKR, and C57BL/6 mice by histology, immunohistochemistry, and flow cytometry. Gastric acid secretion was measured in ligated stomachs, while anti-parietal cell antibodies were assayed by immunofluorescence and enzyme-linked immunosorbent spot assay. SAMP1/YitFcs mice display a corpus-dominant, chronic gastritis with multifocal aggregates of mononuclear cells consisting of T and B lymphocytes. Relatively few aggregates were observed elsewhere in the stomach. The infiltrates in the oxyntic mucosa were associated with the loss of parietal cell mass. AKR mice, the founder strain of the SAMP1/YitFcs, also have gastritis, although they do not develop ileitis. Genetic studies using SAMP1/YitFcs-C57BL/6 congenic mice showed that the genetic regions regulating ileitis had comparable effects on gastritis. The majority of the cells in the aggregates expressed the T cell marker CD3 or the B cell marker B220. Adoptive transfer of SAMP1/YitFcs CD4+ T helper cells, with or without B cells, into immunodeficient recipients induced a pangastritis and duodenitis. SAMP1/YitFcs and AKR mice manifest hypochlorhydria and anti-parietal cell antibodies. These data suggest that common genetic factors controlling gastroenteric disease in SAMP1/YitFcs mice regulate distinct pathogenic mechanisms causing inflammation in separate sites within the digestive tract. PMID:21921286

  9. Modulation of cellular phosphoprotein profiles in transformation and redifferentiation of murine and embryonic fibroblastic cells

    International Nuclear Information System (INIS)

    Chakrabarty, Subhas; Brattain, M.G.

    1985-01-01

    Cellular phosphoprotein profiles from normal mouse embryonic fibroblast AKR-2B cells were compared to those of their permanently, chemically transformed malignant counterparts AKR-MCA cells, and AKR-2B cells reversibly transformed by transforming growth factor (AKR-TGF). Similar 32 P-phosphorylation profiles were observed for both the AKR-TGF and AKR-MCA cells which were distinct from that of the normal AKR-2B cells. Dimethylformamide (DMF)-induced differentiation of the AKR-MCA cells resulted in restoration of the normal AKR-2B phosphorylation profile to the malignant AKR-MCA cells. (author)

  10. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  11. Novel Aldo-Keto Reductases for the Biocatalytic Conversion of 3-Hydroxybutanal to 1,3-Butanediol: Structural and Biochemical Studies

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeho; Flick, Robert; Brunzelle, Joseph; Singer, Alex; Evdokimova, Elena; Brown, Greg; Joo, Jeong Chan; Minasov, George A.; Anderson, Wayne F.; Mahadevan, Radhakrishnan; Savchenko, Alexei; Yakunin, Alexander F. (KRICT); (Toronto); (NWU)

    2017-01-27

    The nonnatural alcohol 1,3-butanediol (1,3-BDO) is a valuable building block for the synthesis of various polymers. One of the potential pathways for the biosynthesis of 1,3-BDO includes the biotransformation of acetaldehyde to 1,3-BDO via 3-hydroxybutanal (3-HB) using aldolases and aldo-keto reductases (AKRs). This pathway requires an AKR selective for 3-HB, but inactive toward acetaldehyde, so it can be used for one-pot synthesis. In this work, we screened more than 20 purified uncharacterized AKRs for 3-HB reduction and identified 10 enzymes with significant activity and nine proteins with detectable activity. PA1127 fromPseudomonas aeruginosashowed the highest activity and was selected for comparative studies with STM2406 fromSalmonella entericaserovar Typhimurium, for which we have determined the crystal structure. Both AKRs used NADPH as a cofactor, reduced a broad range of aldehydes, and showed low activities toward acetaldehyde. The crystal structures of STM2406 in complex with cacodylate or NADPH revealed the active site with bound molecules of a substrate mimic or cofactor. Site-directed mutagenesis of STM2406 and PA1127 identified the key residues important for the activity against 3-HB and aromatic aldehydes, which include the residues of the substrate-binding pocket and C-terminal loop. Our results revealed that the replacement of the STM2406 Asn65 by Met enhanced the activity and the affinity of this protein toward 3-HB, resulting in a 7-fold increase inkcat/Km. Our work provides further insights into the molecular mechanisms of the substrate selectivity of AKRs and for the rational design of these enzymes toward new substrates.

    IMPORTANCEIn this study, we identified several aldo-keto reductases with significant activity in reducing 3

  12. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  13. Evaporation rate measurement in the pool of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

    2000-01-01

    The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

  14. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  15. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  16. Limiting dilution analysis for precursor frequency of Con A-responsive mouse Thy-1+ dendritic epidermal cells

    International Nuclear Information System (INIS)

    Takashima, A.; Bergstresser, P.R.; Nixon-Fulton, J.L.; Tigelaar, R.E.

    1986-01-01

    The authors have recently demonstrated in vitro proliferation of mouse Thy-1 + dendritic epidermal cells (EC) to Con A and IL-2. The purpose of the present study was to utilize limiting dilution analysis to determine the precursor frequency (PF) of Con A-responsive cells within EC enriched by Isolymph centrifugation for Thy-1 + cells (IEC). AKR IEC were cultured in 96 well U-plates (25-75 cells/well) with 2 μg/ml Con A and 2 x 10 5 irradiated (1600 R) AKR spleen cells/well. Cultures were harvested after 7-21 days following 3 H-thymidine pulsing. Results indicated a PF within IEC of 1.5-4.5%. Inclusion of 10 U/ml IL-2 enhanced significantly the proliferation in positive wells but did not alter this PF. In AKR mice, monoclonal antibody 20-10-5S has been shown to react with Thy-1 + EC, but not with peripheral T cells. FACS purification of IEC using 20-10-5S indicated that Con A responsiveness resides exclusively within the 20-10-5S + population. The PF of Con A-responsive Thy-1 + EC was calculated by dividing the PF of IEC by the fraction of 20-10-5S + cells (13-30%) in the IEC suspension. A significant proportion of Thy-1 + EC (∼12%) were found to possess Con A proliferative capacity. These studies will facilitate analysis at a clonal level of possible functional and phenotypic heterogeneity within the Thy-1 + EC population

  17. Molecular characterization of an aldo-keto reductase from Marivirga tractuosa that converts retinal to retinol.

    Science.gov (United States)

    Hong, Seung-Hye; Nam, Hyun-Koo; Kim, Kyoung-Rok; Kim, Seon-Won; Oh, Deok-Kun

    2014-01-01

    A recombinant aldo-keto reductase (AKR) from Marivirga tractuosa was purified with a specific activity of 0.32unitml(-1) for all-trans-retinal with a 72kDa dimer. The enzyme had substrate specificity for aldehydes but not for alcohols, carbonyls, or monosaccharides. The enzyme turnover was the highest for benzaldehyde (kcat=446min(-1)), whereas the affinity and catalytic efficiency were the highest for all-trans-retinal (Km=48μM, kcat/Km=427mM(-1)min(-1)) among the tested substrates. The optimal reaction conditions for the production of all-trans-retinol from all-trans-retinal by M. tractuosa AKR were pH 7.5, 30°C, 5% (v/v) methanol, 1% (w/v) hydroquinone, 10mM NADPH, 1710mgl(-1) all-trans-retinal, and 3unitml(-1) enzyme. Under these optimized conditions, the enzyme produced 1090mgml(-1) all-trans-retinol, with a conversion yield of 64% (w/w) and a volumetric productivity of 818mgl(-1)h(-1). AKR from M. tractuosa showed no activity for all-trans-retinol using NADP(+) as a cofactor, whereas human AKR exhibited activity. When the cofactor-binding residues (Ala158, Lys212, and Gln270) of M. tractuosa AKR were changed to the corresponding residues of human AKR (Ser160, Pro212, and Glu272), the A158S and Q270E variants exhibited activity for all-trans-retinol. Thus, amino acids at positions 158 and 270 of M. tractuosa AKR are determinant residues of the activity for all-trans-retinol. Crown Copyright © 2013. Published by Elsevier B.V. All rights reserved.

  18. Xenotropic type C virus expression in murine thymomas induced by radiation or 3-methylcholanthrene

    International Nuclear Information System (INIS)

    Mayer, A.; Duran-Reynals, M.L.

    1981-01-01

    Thymic lymphoma incidence and thymic expression of MuLV with xenotropic infectivity was monitored in AKR, RF, and reciprocal F 1 mice of the AKR X RF cross after treatment with either γ radiation or the chemical carcinogen 3-methylcholanthrene (MCA). These two inbred strains and the F 1 hybrids developed similary high incidences of thymoma, and lymphomatous cells from AKR mice and (ARK] X RF∫)F 1 mice were observed to be expressing MuLV with xenotropic host range. However, lymphoma cells from RF mice and (RF] X AKR∫)F 1 mice did not shed xenotropic MuLV. Thymic xenotropic virus expression was therefore not correlated with a high incidence of radiation or chemically induced thymoma, but rather appeared to be a phenotype genetically transmitted by AKR mice to F 1 mice of the AKR X RF cross as a dominant trait in induced thymomas. In addition, a maternal effect on thymic xenotropic virus expression in induced thymomas was observed by the comparison of reciprocal F 1 hybrids in this cross

  19. Palmitoylation of the Cysteine Residue in the DHHC Motif of a Palmitoyl Transferase Mediates Ca2+ Homeostasis in Aspergillus.

    Directory of Open Access Journals (Sweden)

    Yuanwei Zhang

    2016-04-01

    Full Text Available Finely tuned changes in cytosolic free calcium ([Ca2+]c mediate numerous intracellular functions resulting in the activation or inactivation of a series of target proteins. Palmitoylation is a reversible post-translational modification involved in membrane protein trafficking between membranes and in their functional modulation. However, studies on the relationship between palmitoylation and calcium signaling have been limited. Here, we demonstrate that the yeast palmitoyl transferase ScAkr1p homolog, AkrA in Aspergillus nidulans, regulates [Ca2+]c homeostasis. Deletion of akrA showed marked defects in hyphal growth and conidiation under low calcium conditions which were similar to the effects of deleting components of the high-affinity calcium uptake system (HACS. The [Ca2+]c dynamics in living cells expressing the calcium reporter aequorin in different akrA mutant backgrounds were defective in their [Ca2+]c responses to high extracellular Ca2+ stress or drugs that cause ER or plasma membrane stress. All of these effects on the [Ca2+]c responses mediated by AkrA were closely associated with the cysteine residue of the AkrA DHHC motif, which is required for palmitoylation by AkrA. Using the acyl-biotin exchange chemistry assay combined with proteomic mass spectrometry, we identified protein substrates palmitoylated by AkrA including two new putative P-type ATPases (Pmc1 and Spf1 homologs, a putative proton V-type proton ATPase (Vma5 homolog and three putative proteins in A. nidulans, the transcripts of which have previously been shown to be induced by extracellular calcium stress in a CrzA-dependent manner. Thus, our findings provide strong evidence that the AkrA protein regulates [Ca2+]c homeostasis by palmitoylating these protein candidates and give new insights the role of palmitoylation in the regulation of calcium-mediated responses to extracellular, ER or plasma membrane stress.

  20. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  1. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  2. Spent fuel management - two alternatives at the FiR 1 reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S.E.J.

    2001-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

  3. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  4. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  5. Properties of autoregressive model in reactor noise analysis, 1

    International Nuclear Information System (INIS)

    Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

    1987-01-01

    Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

  6. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  7. Spent fuel management plans for the FiR 1 Reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S. E. J.

    2002-01-01

    The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

  8. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    International Nuclear Information System (INIS)

    Mesquita, Henrique F.A.; Ferreira, Andrea V.

    2015-01-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  9. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  10. RA reactor operation and maintenance in 1996, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1996-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  11. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  12. The AMPS 1.5 MW low-pressure compact reactor

    International Nuclear Information System (INIS)

    Hewitt, J.S.

    1987-01-01

    The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

  13. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  14. Hic-5’s Regulatory Role in TGFB Signaling in Prostate Stroma

    Science.gov (United States)

    2012-06-01

    the androgen metabolites 3α-Adiol and 3β-Adiol, and their importance is underscored by high expression levels of the aldo keto reductase (AKR1C...known as aldo -keto reductases (AKR1C) [33]. DU145 cells express AKR1C enzymes and are capable of catalyzing redox reactions at the C17 position of...584-95. 37. Bauman, D.R., et al., Development of nonsteroidal anti-inflammatory drug analogs and steroid carboxylates selective for human aldo -keto

  15. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  16. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    Science.gov (United States)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  17. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    International Nuclear Information System (INIS)

    El-genk, M.S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

  18. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  19. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  20. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

    Energy Technology Data Exchange (ETDEWEB)

    Forlerer, Elena; Palacios, Tulio A [comps.

    1998-07-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.

  1. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  2. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  3. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  4. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  5. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  6. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  7. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

    1986-01-01

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

  8. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  9. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  10. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  11. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  12. Dose measurements in controlled area of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, F.L.; Junior, F.M.

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

  13. Properties of auroral kilometric radiation from an interferometer analysis of the ISEE-1 and -2 plasma wave data

    International Nuclear Information System (INIS)

    Baumback, M.M.

    1986-01-01

    The first satellite-satellite interferometery measurements of the auroral kilometric radiation (AKR) source region diameter are presented. By correlating the analog waveforms detected by ISEE-1 and ISEE-2, the size of the AKR source region is determined. Correlations have been measured at 125 and 250 kHz for projected baselines ranging from 20 to 3868 km. High correlations are found at all projected baselines, with little or no tendency to decrease at long baselines. The correlation is lower for events with wide bandwidths than for events with narrow bandwidths. The magnitude of the correlation as a function of signal delay and the spectra of the individual bursts show that sometimes the bandwidth of AKR bursts varies rapidly and can be narrower than 20 Hz. The spectra observed by both spacecraft are nearly identical. Correlation results are interpreted differently for incoherent radiation than for coherent radiation. If the radiation is incoherent, the visibility of the source region is the Fourier transform of the brightness distribution. Assuming incoherent radiation the average source region diameter for all analyzed bursts is less than 9.27 km. Source region diameters measured for individual bursts range from 1 to 16 km. Generation mechanisms that only amplify incoming radiation cannot produce high correlations unless the source region diameter is smaller than 25 km

  14. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  15. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  16. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    International Nuclear Information System (INIS)

    Olsson, S.; Andermo, L.

    1980-01-01

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  17. High dose rate brachytherapy source measurement intercomparison.

    Science.gov (United States)

    Poder, Joel; Smith, Ryan L; Shelton, Nikki; Whitaker, May; Butler, Duncan; Haworth, Annette

    2017-06-01

    This work presents a comparison of air kerma rate (AKR) measurements performed by multiple radiotherapy centres for a single HDR 192 Ir source. Two separate groups (consisting of 15 centres) performed AKR measurements at one of two host centres in Australia. Each group travelled to one of the host centres and measured the AKR of a single 192 Ir source using their own equipment and local protocols. Results were compared to the 192 Ir source calibration certificate provided by the manufacturer by means of a ratio of measured to certified AKR. The comparisons showed remarkably consistent results with the maximum deviation in measurement from the decay-corrected source certificate value being 1.1%. The maximum percentage difference between any two measurements was less than 2%. The comparisons demonstrated the consistency of well-chambers used for 192 Ir AKR measurements in Australia, despite the lack of a local calibration service, and served as a valuable focal point for the exchange of ideas and dosimetry methods.

  18. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  19. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  20. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  1. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  2. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  3. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  4. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  5. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  6. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  7. Correlations of auroral kilometric radiation with Birkeland currents

    International Nuclear Information System (INIS)

    Saflekos, N.A.; Carovillano, R.L.; Sheehan, R.E.

    1983-01-01

    This chapter examines auroral kilometric radiation (AKR) in relation to the strength of field-aligned currents (FAC), which represent an energy source stored in the form of magnetic field energy density in the neighborhood of the earth. An attempt is made to find a direct relationship between AKR power flux and optical auroral emissions. Topics considered include correlated Hawkeye and Triad satellite observations and correlated AKR and optical emissions. It is indicated that AKR is electromagnetic radiation in the frequency range of 50 to 500 kHz; AKR is generated at frequencies above the electron plasma frequency and below the electron gyrofrequency; AKR propagates in the Right Hand Extraordinary mode; and AKR may show fine structure in frequency. The principal findings include: distributions of AKR intensity with increasing auroral activity show that although quiet and disturbed auroras are generally accompanied by weak and intense AKR, the moderate auroras are associated with a broad range of AKR power; distributions of AKR intensity with increasing auroral electrojet (AE) index during the expansion phase of a polar magnetic substorm show near maximum levels of AKR power emission; and the maximum AKR power increases with increasing auroral activity and with increasing Birkeland current strength

  8. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    1988-03-01

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  9. Transformation of 1,1,1-trichloroethane in an anaerobic packed-bed reactor at various concentrations of 1,1,1-trichloroethane, acetate and sulfate

    NARCIS (Netherlands)

    deBest, JH; Jongema, H; Weijling, A; Doddema, HJ; Janssen, DB; Harder, W

    Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.

  10. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  11. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

  12. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  13. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  14. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  15. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  16. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  17. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

  18. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

  19. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  20. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  1. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  2. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  3. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  4. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  5. An Epidemiologic Study of Genetic Variation in Hormonal Pathways in Relation to the Effect of Hormone Replacement Therapy on Breast Cancer Risk

    Science.gov (United States)

    2008-10-01

    PGR gene to be associated with breast cancer. Using long-range PCR techniques to sequence exons 1 and 2 of PGR, and a Solexa chip from Illumina...specific histologic types associated with single SNPs in PGR, AKR1C1, AKR1C2, AKR1C3, SRD5A1, SRD5A2 and CYP3A4 Breast caner overall Ductal Lobular...1.3 0.96 T/T 92 (9.1) 120 (9.6) 1.1 0.8 1.4 72 (9.5) 1.1 0.8 1.5 27 (9.7) 1.0 0.6 1.7 0.93 CYP3A4 rs12333983 T/T 791 (77.8) 985

  6. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  7. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  8. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  9. Correlation between Auroral kilometric radiation and field-aligned currents

    International Nuclear Information System (INIS)

    Green, J.L.; Saflekos, N.A.; Gurnett, D.A.; Potemra, T.A.

    1982-01-01

    Simultaneous observations of field-aligned currents (FAC) and auroral kilometric radiation (AKR) are compared from the polar-orbiting satellites Triad and Hawkeye. The Triad observations were restricted to the evening-to-midnight local time sector (1900 to 0100 hours magnetic local time) in the northern hemisphere. This is the region in which the most intense storms of AKR are believed to originate. The Hawkeye observations were restricted to when the satellite was in the AKR emission cone in the northern hemisphere and at radial distances > or =7R/sub E/ (earth radii) to avoid local propagation cutoff effects. A(R/7R/sub E/) 2 normalization to the power flux measurements of the kilometric radiation from Hawkeye is used to take into account the radial dependence of this radiation and to scale all intensity measurements so that they are independent of Hawkeye's position in the emission cone. Integrated field-aligned current intensities from Triad are determined from the observed transverse magnetic field disturbances. There appears to be a weak correlation between AKR intensity and the integrated current sheet intensity of field-aligned currents. In general, as the intensity of auroral kilometric radiation increases so does the integrated auroral zone current sheet intensity increase. Statistically, the linear correlation coefficient between the log of the AKR power flux and the log of the current sheet intensity is 0.57. During weak AKR bursts ( - 18 W m - 2 Hz - 1 ), Triad always observed weak FAC'S ( - 1 ), and when Triad observed large FAC's (> or =0.6 A m - 1 ), the AKR intensity from Hawkeye was moderately intense (10 - 5 to 10 - 14 W m - 2 Hz - 1 ) to intense (>10 - 14 W m - 2 Hz - 1 ). It is not clear from these preliminary results what the exact role is that auroral zone field-aligned currents play in the generation or amplification of auroral kilometric radiation

  10. Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

    International Nuclear Information System (INIS)

    Tikhomirov, L.N.

    1995-01-01

    The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

  11. RA reactor operation and maintenance in 1994, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1994-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  12. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  13. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  14. [On the role of selective silencer Freud-1 in the regulation of the brain 5-HT(1A) receptor gene expression].

    Science.gov (United States)

    Naumenko, V S; Osipova, D V; Tsybko, A S

    2010-01-01

    Selective 5-HT(1A) receptor silencer (Freud-1) is known to be one of the main factors for transcriptional regulation of brain serotonin 5-HT(1A) receptor. However, there is a lack of data on implication of Freud-1 in the mechanisms underlying genetically determined and experimentally altered 5-HT(1A) receptor system state in vivo. In the present study we have found a difference in the 5-HT(1A) gene expression in the midbrain of AKR and CBA inbred mouse strains. At the same time no distinction in Freud-1 expression was observed. We have revealed 90.3% of homology between mouse and rat 5-HT(1A) receptor DRE-element, whereas there was no difference in DRE-element sequence between AKR and CBA mice. This indicates the absence of differences in Freud-1 binding site in these mouse strains. In the model of 5-HT(1A) receptor desensitization produced by chronic 5-HT(1A) receptor agonist administration, a significant reduction of 5-HT(1A) receptor gene expression together with considerable increase of Freud-1 expression were found. These data allow us to conclude that the selective silencer of 5-HT(1A) receptor, Freud-1, is involved in the compensatory mechanisms that modulate the functional state of brain serotonin system, although it is not the only factor for 5-HT(1A) receptor transcriptional regulation.

  15. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  16. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    1986-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  17. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  18. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  19. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  20. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  1. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

  2. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  3. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  4. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  5. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  6. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2000-01-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

  7. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  8. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  9. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  10. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  11. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  12. Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

    2000-07-01

    It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

  13. The nature of tolerance in adult recipient mice made tolerant of alloantigens with supralethal irradiation followed by syngeneic bone marrow cell transplantation plus injection of F1 spleen cells

    International Nuclear Information System (INIS)

    Tomita, Y.; Himeno, K.; Mayumi, H.; Tokuda, N.; Nomoto, K.

    1989-01-01

    The length of time after syngeneic bone marrow reconstitution when tolerance to alloantigens can be induced in adult mice during T cell differentiation from bone marrow cells was studied by exposing those T cells to (recipient x donor)F1 spleen cells. Supralethally irradiated C3H/He Slc(C3H; H-2k) mice were reconstituted with 1 x 10(7) syngeneic T cell-depleted bone marrow cells and then injected intravenously with 5 x 10(7) (C3H x C57BL/6[B6])F1 (B6C3F1; H-2bxk) or (C3H x AKR/J[AKR])F1 (AKC3F1; H-2kxk) spleen cells at various intervals. In the fully allogeneic combination of B6C3F1----C3H, EL-4 tumor originating from B6 was accepted, and survival of grafted B6 skin was significantly prolonged in the tolerant C3H mice treated with irradiation on day -1 followed by injection of syngeneic bone marrow cells on day 0 plus B6C3F1 spleen cells on days 0, 5, or 10, in a tolerogen-specific manner. In the multiminor histocompatibility antigen-disparate combination of AKC3F1----C3H, AKR skin grafts were permanently accepted in the tolerant C3H mice treated with AKC3F1 spleen cells on days 0, 5, 10, or 15. Immunological parameters, including cytotoxic T lymphocyte activity and delayed foot-pad reaction (DFR), were almost completely suppressed in C3H mice made tolerant of B6 or AKR antigens. A chimeric assay using a direct immunofluorescence method revealed that the tolerant C3H mice given B6C3F1 spleen cells on day 0 were mixed-chimeric for at least 8 weeks after syngeneic bone marrow reconstitution, but not definitely chimeric thereafter. The C3H mice given AKC3F1 spleen cells on day 0 were chimeric even 43 weeks after syngeneic bone marrow reconstitution, but the C3H mice given AKC3F1 spleen cells on day 15 showed temporal chimerism that disappeared within 43 weeks. The untolerant mice were never detectably chimeric

  14. Correlation between auroral kilometric radiation and inverted v electron precipitation

    International Nuclear Information System (INIS)

    Green, J.L.; Gurnfti, D.A.; Hoffmans, R.A.

    1979-01-01

    Simultaneous observations of energetic electron precipitations and auroral kilometric radiation (AKR) were obtained from the polar orbiting satellites AE-D and Hawkeye. The Hawkeye observations were restricted to periods when the satellite was in the AKR emission cone in the northern hemisphere an at radial distances > or approx. =7 R/sub E/ to avoid local propagation cutoff effects. In addition, the AE-D measurements were restricted to complete passes across the auroral oval in the evening to midnight local time sector (from 20 to 01 hours magnetic local time). This is the local time region where the most intense bursts of AKR are believed to originate. A qualitative survey of AKR and electron precipitation than with plasma sheet precipitation. Quantitatively, a good correlation is found between the AKR intensity and the peak energy of inverted V events. In addition, in the tail of the most field-aligned portion (approx.O 0 pitch angle) of the distribution functions of the inverted V events,systematic changes are indicated as the associated AKR intensity increases. When the AKR power flux is weak ( -17 W/(m 2 Hz)). From a determination of the simultaneous power in the inverted V events and the AKR bursts, the efficiency of converting the charge particle energy into EM radiation increases to a maximum of about 1% for the most intense AKR bursts. However, conversion efficiencies as low as 10 -5 % are also found. There is some evidence which suggests that the tail temperature, T in F (V) of the inverted V events, may play an important role in the efficient generation or amplification of auroral kilometric radiation

  15. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  16. Applied research into direct numerical control of A-1 reactor temperature

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.

    1974-01-01

    Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

  17. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  18. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  19. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  20. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  1. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  2. NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Paiva, Rosemeire P.; Salvetti, Tereza C.

    2005-01-01

    Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

  3. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  4. Reactor calculations in aid of isotope production at SAFARI-1

    International Nuclear Information System (INIS)

    Ball, G.

    2003-01-01

    Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

  5. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  6. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

  7. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carvalho, Ricardo Pinto de

    2006-01-01

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  8. German private mobile telecommunication research program. Effects of a) low-frequency magnetic fields, b) GSM mobile telecommunication signals, c) UMTS signals on the spontaneous leukemia rate of AKR mice

    International Nuclear Information System (INIS)

    2006-01-01

    Since the introduction of private mobile telecommunication about 10 years ago, there is a worldwide discussion about the possible adverse health effects of such highfrequency electromagnetic fields (EMF). Main concerns are regarding the possible cancerogenic or promoting effect of mobile phones and base stations. The insecurity of the public is increased by the continuous development of new techniques and introduction of new systems (e.g. wireless LAN, UMTS), for which insufficient data are available about their possible health hazards. This project's aim was therefore to investigate if chronic exposure to EMF of the UMTS-standard will influence leukaemia or the development of solid tumors in an animal model. The experiment was performed with AKR mice, which is an accepted animal model for leukaemia. With high probability, these mice develop the disease within their lifespan, so that the time course of the onset of leukaemia, bodyweight, blood picture, time of survival, and pathological examinations were the endpoints of the investigation. 160 female AKR/J mice were exposed or sham-exposed for 24 hours per day to EMF of the UMTS-standard (0.4 W/kg SAR). Additionally, 33 animals were kept as cagecontrols. For this experiment special radial waveguides were constructed and built which have a low SAR variability. All experiments were performed in a blinded fashion. The only significant difference between exposed and sham-exposed mice was a higher survival rate of exposed animals. However, the number of ill animals, the mean survival time and the severity code of the disease did not differ between the experimental groups. Animals of the cage-control group had a significant lower growth rate compared to animals kept in the radial waveguides, likely the cause of differences in the method of food supply. The results of this study don't show any negative effects of the exposure. The results do not show any negative effects of exposure for months to EMF at levels 5-fold the

  9. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    Klein, J.; Medel, J.; Daie, J.; Torres, H.

    2000-01-01

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  10. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  11. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  12. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  13. An economic analysis of stretch-out for Angra-1 reactor

    International Nuclear Information System (INIS)

    Sakai, M.

    1989-01-01

    An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)

  14. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  15. Welding electrode for peripheral welds of A-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    1975-01-01

    The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

  16. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  17. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  18. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  19. . Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

    International Nuclear Information System (INIS)

    Yusuf, I; Mati, A. A.

    2010-01-01

    The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

  20. VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.

    2003-01-01

    The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

  1. FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    Auterinen, I.; Salmenhaara, S.E.J. . Author

    2004-01-01

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

  2. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  3. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  4. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  5. Nuclear material control at IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    1988-01-01

    The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

  6. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  7. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits

  8. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  9. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  10. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  11. 1DB, a one-dimensional diffusion code for nuclear reactor analysis

    International Nuclear Information System (INIS)

    Little, W.W. Jr.

    1991-09-01

    1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

  12. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  13. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  14. RA reactor operation and maintenance in 1989, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Sanovic, V.

    1989-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  15. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  16. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  17. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  18. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  19. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  1. Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

  2. Observations pertaining to the generation of auroral kilometric radiation

    International Nuclear Information System (INIS)

    Green, J.L.

    1981-01-01

    Auroral kilometric radiation (AKR) observations that have determined the propagation mode or polarization of the radiation and the detailed intensity distribution of the AKR emission cone are discussed. Attention is also given to correlations between AKR with discrete field-aligned currents. It is noted that these observations have helped to identify the auroral particle population most likely responsible for the generation of AKR and the possible sources of the free energy that drives the instability. Thus far, AKR has not been observed simultaneously with large electrostatic waves. Auroral zone current systems are thought to be intimately involved in the generation of AKR. In particular, the most probable source of energy for AKR is the precipitating inverted-V auroral electron distribution

  3. Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

    2014-08-15

    IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

  4. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  5. Operating reactors licensing actions summary. Volume 5, Number 1

    International Nuclear Information System (INIS)

    1985-03-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  6. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  7. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  8. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  9. Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Perkins, H.K.

    1977-08-01

    Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

  10. Source region of aurora kilometric radiation

    International Nuclear Information System (INIS)

    Morioka, Akira; Oya, Hiroshi; Tokumaru, Munetoshi

    1981-01-01

    This paper discusses the source region of aurora kilometric radiation (AKR), and the relation between the particle acceleration region and the polar ionosphere. The observation was made by the satellite 'Jikiken'. The AKR can be transferred to Jikiken without any interception, when the magnetic latitude of the apogee of the satellite is low. The spectra taken in June, 1980, were analyzed. The observed spectra showed the source regions of the AKR were in the aurora bands of the north and south poles. One example showed that the 200 kHz component of AKR from both poles showed the similar behavior, and another example showed that the AKR spectra from both poles showed different behavior. The altitude distribution of source regions was able to be obtained. The altitude of AKR-A was in the range between 6200 and 12000 km, and that of AKR-B was in the range of 3500 and 5200 km. The source of AKR-A was identified as that in the south hemisphere, and that of AKR-B in the north hemisphere. The asymmetric spectra of AKR-A and B showed that the spread and intensity of the electric field along magnetic lines generated above the polar ionosphere were related with the conditions of the ionosphere. (Kato, T.)

  11. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  12. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  13. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  14. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  15. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  16. Restricted antibody formation to sheep erythrocytes of allogeneic bone marrow chimeras histoincompatible at the K end of the H-2 complex

    International Nuclear Information System (INIS)

    Onoe, K.; Yasumizu, R.; Oh-Ishi, T.; Kakinuma, M.; Good, R.A.; Morikawa, K.

    1981-01-01

    Employing a new method for allogeneic bone marrow transplantation, irradiation chimeras constructed from various combinations of marrow cells from B10 H-2 recombinant mice and AKR recipients were prepared. Though these chimeras had well-developed populations of T and B cells, they showed strikingly different patterns of responses in the primary antibody formation to sheep erythrocytes (SRBC), a T dependent antigen. These are (a) AKR mice treated with C57BL/10 cells, [B10 leads to AKR] fully H-2 incompatible, and AKR mice treated with B10.A (5R) cells, [5R leads to AKR] I-J,E compatible chimeras that were almost completely unresponsive to SRBC; (b) AKR mice treated with B10.BR cells [BR leads to AKR] fully H-2 compatible, and AKR mice treated with B10 AKM cells, [AKM leads to AKR] chimeras where donor and recipient differed only at H-2D, showed the same number of plaque-forming cells (PFC) as B10 control mice; (c) AKR mice treated with B10.A cells, [B10 leads to AKR] chimeras, where donor and recipient were matched at H-2K-I-E region, showed about one-half the number of PFC as the control mice. From these results we conclude that in allogeneic bone marrow chimeras primary antibody response to T-dependent antigen, such as SRBC, is generated when at least the K end of the H-2 complex is compatible between donor and recipient

  17. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

  18. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  19. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  20. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  1. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  2. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  3. Isoenzyme-specific up-regulation of glutathione transferase and aldo-keto reductase mRNA expression by dietary quercetin in rat liver.

    Science.gov (United States)

    Odbayar, Tseye-Oidov; Kimura, Toshinori; Tsushida, Tojiro; Ide, Takashi

    2009-05-01

    The impact of quercetin on the mRNA expression of hepatic enzymes involved in drug metabolism was evaluated with a DNA microarray and real-time PCR. Male Sprague-Dawley rats were fed an experimental diet containing either 0, 2.5, 5, 10, or 20 g/kg of quercetin for 15 days. The DNA microarray analysis of the gene expression profile in pooled RNA samples from rats fed diets containing 0, 5, and 20 g/kg of quercetin revealed genes of some isoenzymes of glutathione transferase (Gst) and aldo-keto reductase (Akr) to be activated by this flavonoid. Real-time PCR conducted with RNA samples from individual rats fed varying amounts of quercetin together with the microarray analysis showed that quercetin caused marked dose-dependent increases in the mRNA expression of Gsta3, Gstp1, and Gstt3. Some moderate increases were also noted in the mRNA expression of isoenzymes belonging to the Gstm class. Quercetin also dose-dependently increased the mRNA expression of Akr1b8 and Akr7a3. However, it did not affect the parameters of the other Gst and Akr isoenzymes. It is apparent that quercetin increases the mRNA expression of Gst and Akr involved in drug metabolism in an isoenzyme-specific manner. Inasmuch as Gst and Akr isoenzymes up-regulated in their gene expression are involved in the prevention and attenuation of cancer development, this consequence may account for the chemopreventive propensity of quercetin.

  4. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  5. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  6. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  7. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  8. Substrate Specificity, Inhibitor Selectivity and Structure-Function Relationships of Aldo-Keto Reductase 1B15: A Novel Human Retinaldehyde Reductase

    Czech Academy of Sciences Publication Activity Database

    Giménez-Dejoz, J.; Kolář, Michal H.; Ruiz, F. X.; Crespo, I.; Cousido-Siah, A.; Podjarny, A.; Barski, O. A.; Fanfrlík, Jindřich; Parés, X.; Farrés, J.; Porté, S.

    2015-01-01

    Roč. 10, č. 7 (2015), e0134506/1-e0134506/19 E-ISSN 1932-6203 Institutional support: RVO:61388963 Keywords : retinoic acid biosynthesis * site-directed mutagenesis * tumor marker AKR1B15 Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 3.057, year: 2015 http://journals.plos.org/plosone/article?id=10.1371/journal.pone.0134506

  9. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Maynard, C.W.

    1984-04-01

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  10. PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

    1990-06-01

    Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

  11. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  12. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  13. Neutron radiography in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

    1988-08-01

    Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

  14. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  15. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  16. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

    2014-10-01

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  17. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  18. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  19. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  20. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  1. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  2. Measurements of reactivity of reactor G1

    International Nuclear Information System (INIS)

    Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

    1957-01-01

    The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr

  3. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    Larry B. Wimmer

    2001-01-01

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)

  4. Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

    International Nuclear Information System (INIS)

    Moreira, Joao Manoel Losada; Frajndlich, Roberto

    1995-01-01

    This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

  5. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

    2017-08-30

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO3-xPbTiO3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

  6. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  7. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...

  8. Reactor of the XXI century

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.; Solov'ev, Yu.A.

    2001-01-01

    The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

  9. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  10. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  11. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  12. Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

    International Nuclear Information System (INIS)

    Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

    2000-01-01

    The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

  13. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  14. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  15. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  16. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  17. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  18. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

    2005-01-01

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  19. Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

  20. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  1. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  2. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  3. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  4. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  5. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1973-01-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  6. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  7. Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-05-01

    During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity

  8. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  9. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  10. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  11. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  12. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

    1986-01-01

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

  13. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  14. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  15. Pre-clinical activity of PR-104 as monotherapy and in combination with sorafenib in hepatocellular carcinoma.

    Science.gov (United States)

    Abbattista, Maria R; Jamieson, Stephen M F; Gu, Yongchuan; Nickel, Jennifer E; Pullen, Susan M; Patterson, Adam V; Wilson, William R; Guise, Christopher P

    2015-01-01

    PR-104 is a clinical stage bioreductive prodrug that is converted in vivo to its cognate alcohol, PR-104A. This dinitrobenzamide mustard is reduced to activated DNA cross-linking metabolites (hydroxylamine PR-104H and amine PR-104M) under hypoxia by one-electron reductases and independently of hypoxia by the 2-electron reductase aldo-keto reductase 1C3 (AKR1C3). High expression of AKR1C3, along with extensive hypoxia, suggested the potential of PR-104 for treatment of hepatocellular carcinoma (HCC). However, a phase IB trial with sorafenib demonstrated significant toxicity that was ascribed in part to reduced PR-104A clearance, likely reflecting compromised glucuronidation in patients with advanced HCC. Here, we evaluate the activity of PR-104 in HCC xenografts (HepG2, PLC/PRF/5, SNU-398, Hep3B) in mice, which do not significantly glucuronidate PR-104A. Cell line differences in sensitivity to PR-104A in vitro under aerobic conditions could be accounted for by differences in both expression of AKR1C3 (high in HepG2 and PLC/PRF/5) and sensitivity to the major active metabolite PR-104H, to which PLC/PRF/5 was relatively resistant, while hypoxic selectivity of PR-104A cytotoxicity and reductive metabolism was greatest in the low-AKR1C3 SNU-398 and Hep3B lines. Expression of AKR1C3 in HepG2 and PLC/PRF/5 xenografts was in the range seen in 21 human HCC specimens. PR-104 monotherapy elicited significant reductions in growth of Hep3B and HepG2 xenografts, and the combination with sorafenib was significantly active in all 4 xenograft models. The results suggest that better-tolerated analogs of PR-104, without a glucuronidation liability, may have the potential to exploit AKR1C3 and/or hypoxia in HCC in humans.

  16. Experimental radioimmunotherapy with I-131-antibody against a differentiation antigen

    International Nuclear Information System (INIS)

    Badger, C.C.; Krohn, K.A.; Bernstein, I.D.

    1985-01-01

    The authors have previously shown that I-131-labeled antibodies (Ab) against the Thyl.l antigen can care AKR/Cu (Thyl.2+) mice bearing the AKR/J (Thy 1l.1+) SL2 T-cell lymphoma. The authors have now extended these studies to therapy with I-131-anti-Thyl.1 of SL2 lymphoma in AKR/J mice where Ab reacts with both tumor and normal cells. A 25 μg bolus was rapidly cleared from serum by binding to spleen cells (75% with Tl/2 <60 min.) and only low concentrations of Ab(<2% ID/gm) were present in tumor after infusion. Therapy of AKR/J mice bearing established s.c. lymphoma nodules with 1500 μCi I-131-anti-Thyl.1 resulted in complete regression of the nodule in 6/6 animals although tumor eventually regrew and all animals died of metastatic lymphoma. In contrast, I-131-irrelevant Ab given to produce the same amount of whole body radiation (750 μCi) did not affect tumor growth. These studies suggest that radiolabeled-AB against differentiation antigens may be useful for therapy in spite of binding to normal cell populations

  17. Generation of auroral kilometric radiation and the structure of auroral acceleration region

    International Nuclear Information System (INIS)

    Lee, L.C.; Kan, J.R.; Wu, C.S.

    1980-01-01

    Generation of auroral kilometric radiation (AKR) in the auroral acceleration region is studied. It is shown that auroral kilometric radiation can be generated by the backscattered electrons trapped in the acceleration region via a cyclotron maser process. The parallel electric field in the acceleration region is required to be distributed over 1-2 Rsub(E). The observed AKR frequency spectrum can be used to estimate the altitude range of the auroral acceleration region. The altitudes of the lower and upper boundaries of the acceleration region determined from the AKR data are respectively approximately 2000 and approximately 9000 km. (author)

  18. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  19. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

  20. Security devices and experiment facilities at ENEA TRIGA RC-1 reactor

    International Nuclear Information System (INIS)

    Bianchi, P.; Festinesi, A.; Santoro, E.; Tardani, G.; Magli, M.; Reis, G.

    1990-01-01

    RC-1 TRIGA operating exercise staff has produced some auxiliary security devices. These are the neutron source automatic handling device, irradiated samples rabbit connection rotating rack, and auxiliary equipment for transferring hot fuel elements. The reactor electronic control instrumentation system includes various instrumentation channels, the operating capability of which must be verified by the licensee as per Italian regulations. In order to obtain automatic and repeatable operations, TEMAV designed and constructed a remotely-driven source transfer device, based on requirements, performance specifications and technical data supplied by ENEA-TIB. The pneumatic irradiating system for short lived materials allows extraction of radiated samples in a time no longer than 4 seconds. To optimize the system, both as to operability and health protection, a specific rotating rack for the connection of irradiated samples with pneumatic transfer (RABBIT) was produced. To permit 1 MW hot fuel element storage in pits it is necessary to remove hot 100 KW fuel elements and transfer them to a re-treatment plant. Feasibility studies showed the impossibility of using heavy trucks inside the reactor hall. To avoid problems trucks are left outside the reactor hall and only the PEGASO container is removed with a special device that runs on rails. Movement from Rail truck is assured by an electromotor driving pull device and security cable

  1. Qualification process of dispersion fuels in the IEAR1 research reactor

    International Nuclear Information System (INIS)

    Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

  2. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  3. Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp strain DCA1

    NARCIS (Netherlands)

    Hage, J.C.; Houten, R.T.; Tramper, J.; Hartmans, S.

    2004-01-01

    A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown

  4. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  5. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1)

    International Nuclear Information System (INIS)

    Forlerer, Elena; Palacios, Tulio A.

    1998-01-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation

  6. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

  7. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included.

  8. Proposal to the United States Energy Research and Development Administration for continuation of fusion reactor technology studies. Progress report October 1, 1977--July 1, 1978

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1978-01-01

    Since the last progress report we have concentrated on three main areas of research: (1) the study of the NUWMAK reactor design, (2) the study of rf heating for tokamak reactors, and (3) the initiation of a tandem mirror reactor study. The initial work on the tandem mirror reactor is included as background in the technical proposal. Summaries of our work on recent assessments of lithium reserves and neutral transport codes are included

  9. University Reactor Sharing Program. Period covered: September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Hajek, B.K.; Myser, R.D.; Miller, D.W.

    1982-12-01

    During the period from September 1, 1981 to August 31, 1982, the Ohio State University Nuclear Reactor Laboratory participated in the Reactor Sharing Program by providing services to eight colleges and universities. A laboratory on Neutron Activation Analysis was developed for students in the program. A summary of services provided and a copy of the laboratory procedure are attached. Services provided in the last funded period were in three major areas. These were neutron activation analysis, nuclear engineering labs, and introductions to nuclear research. One group also performed radiation surveys and produced isotopes for calibration of their own analytical equipment

  10. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, A. P. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Matos, J. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations of the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.

  11. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  12. Upgrade of Instrumentation for Purdue Reactor PUR-1

    International Nuclear Information System (INIS)

    Revankar, S.T.; Merritt, E.; Bean, R.

    2000-01-01

    The major objective of this program was to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective two projects on instrument upgrade were implemented. The first one was to convert the vacuum tube control and safety amplifiers (CSA) to solid state electronics, and the other was to upgrade the electrical and electronic shielding. This report is the annual report and gives the efforts and progress achieved on these two projects from July 1999 to June 2000

  13. Main refurbishment activities on electronic and electrical equipment for the FRG-1 research reactor

    International Nuclear Information System (INIS)

    Blom, K.H.; Krull, W.

    1997-01-01

    As GKSS intends to operate the research reactor FRG-1 safely and reliably for many years to come, the plant is constantly refurbished and upgraded both in the interests of safety and operational reasons. The following electronic and electrical systems have been replaced or improved since 1990: Information and signalling systems; Emergency power plant (permit applied for); External and internal lightning protection system; Reactor protection system (in part); Safety lighting; Alarm and staff locating system; Control room telephone system; Closed-circuit television system; Beam tube controls; Storage plant for radioactive liquid waste; Ambient dose rate measuring system; Meteorological measuring system; Control and measuring system for the primary cooling circuit; Control rod drives; Control rod control system; Soft start for the secondary pumps; Control and switching devices for the emergency power plant; Trailing cable installation for the reactor bridge; Main-voltage distribution systems/cable routes. (author). 13 figs, 1 tab

  14. Modernization of turbine control system and reactor control system in Almaraz 1 and 2; MOdernizacion de los sistemas de control de turbina y del reactor en Almaraz 1 y 2

    Energy Technology Data Exchange (ETDEWEB)

    Pulido, C.; Diez, J.; Carrasco, J. A.; Lopez, L.

    2005-07-01

    The replacement of the Turbine Control System and Reactor Control System are part of the Almaraz modernization program for the Instrumentation and Control. For these upgrades Almaraz has selected the Ovation Platform that provides open architecture and easy expansion to other systems, these platforms is highly used in many nuclear and thermal plants around the world. One of the main objective for this project were to minimize the impact on the installation and operation of the plant, for that reason the project is implemented in two phases, Turbine Control upgrade and Reactor Control upgrade. Another important objective was to increase the reliability of the control system making them fully fault tolerant to single failures. The turbine Control System has been installed in Units 1 and 2 while the Reactor Control System will be installed in 2006 and 2007 outages. (Author)

  15. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  16. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    Maillard, M.L.

    1960-01-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

  17. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  18. Identification, purification and characterization of furfural transforming enzymes from Clostridium beijerinckii NCIMB 8052.

    Science.gov (United States)

    Zhang, Yan; Ujor, Victor; Wick, Macdonald; Ezeji, Thaddeus Chukwuemeka

    2015-06-01

    Generation of microbial inhibitory compounds such as furfural and 5-hydroxymethylfurfural (HMF) is a formidable roadblock to fermentation of lignocellulose-derived sugars to butanol. Bioabatement offers a cost effective strategy to circumvent this challenge. Although Clostridium beijerinckii NCIMB 8052 can transform 2-3 g/L of furfural and HMF to their less toxic alcohols, higher concentrations present in biomass hydrolysates are intractable to microbial transformation. To delineate the mechanism by which C. beijerinckii detoxifies furfural and HMF, an aldo/keto reductase (AKR) and a short-chain dehydrogenase/reductase (SDR) found to be over-expressed in furfural-challenged cultures of C. beijerinckii were cloned and over-expressed in Escherichia coli Rosetta-gami™ B(DE3)pLysS, and purified by histidine tag-assisted immobilized metal affinity chromatography. Protein gel analysis showed that the molecular weights of purified AKR and SDR are close to the predicted values of 37 kDa and 27 kDa, respectively. While AKR has apparent Km and Vmax values of 32.4 mM and 254.2 mM s(-1) respectively, using furfural as substrate, SDR showed lower Km (26.4 mM) and Vmax (22.6 mM s(-1)) values on the same substrate. However, AKR showed 7.1-fold higher specific activity on furfural than SDR. Further, both AKR and SDR were found to be active on HMF, benzaldehyde, and butyraldehyde. Both enzymes require NADPH as a cofactor for aldehydes reduction. Based on these results, it is proposed that AKR and SDR are involved in the biotransformation of furfural and HMF by C. beijerinckii. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  20. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  1. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  2. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  3. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  4. Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Ricardo, Carla Pereira; Mesquita, Amir Zacarias

    2007-01-01

    During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

  5. The different generation of nuclear reactors from Generation-1 to Generation-4

    International Nuclear Information System (INIS)

    Cognet, G.

    2010-01-01

    In this work author deals with the history of the development of nuclear reactors from Generation-1 to Generation-4. The fuel cycle and radioactive waste management as well as major accidents are presented, too.

  6. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

  7. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  8. Viking observations at the source region of auroral kilometric radiation

    International Nuclear Information System (INIS)

    Bahnsen, A.; Jespersen, M.; Ungstrup, E.; Pedersen, B.M.; Eliasson, L.; Murphree, J.S.; Elphinstone, R.D.; Blomberg, L.; Holmgren, G.; Zanetti, L.J.

    1989-01-01

    The orbit of the Swedish satellite Viking was optimized for in situ observations of auroral particle acceleration and related phenomena. In a large number of the orbits, auroral kilometric radiation (AKR) was observed, and in approximately 35 orbits the satellite passed through AKR source regions as evidenced by very strong signals at the local electron cyclotron frequency f ce . These sources were found at the poleward edge of the auroral oval at altitudes, from 5,000 to 8,000 km, predominantly in the evening sector. The strong AKR signal has a sharp low-frequency cutoff at or very close to f ce in the source. In addition to AKR, strong broadband electrostatic noise is measured during the source crossings. Energetic (1-15 keV) electrons are always present at and around the AKR sources. Upward directed ion beams of several keV are closely correlated with the source as are strong and variable electric fields, indicating that a region of upward pointing electric field below the observation point is a necessary condition for AKR generation. The plasma density is measured by three independent experiments and it is generally found that the density is low across the whole auroral oval. For some source crossings the three methods agree and show a density depletion (but not always confined to the source region itself), but in many cases the three measurements do not yield consistent results. The magnetic projection of the satellite passes through auroral forms during the source crossings, and the strongest AKR events seem to be connected with kinks in an arc or more complicated structures

  9. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  10. Feasibility study of application of Prompt Gamma Neutron Activation Analysis (PGNAA) method in TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Guerra, Bruno Teixeira

    2016-01-01

    The TRIGA Mark I IPR-R1 research reactor is located at Nuclear Technology Development Centre (CDTN), Brazilian Commission for Nuclear Energy (CNEN), in Belo Horizonte, Brazil. The reactor operates at 100 kW but the core configuration allows the increasing of the power up to 250 kW. It has been applied research, training and radioisotopes production. The establishment of the Prompt Gamma Neutron Activation Analysis (PGNAA) method at the TRIGA IPR-R1 reactor will significantly increase the types of matrices analysed as well as the number of chemical elements. Additionally it will complement the neutron activation analysis. This work presents a proposed design of a PGNAA facility to be installed at the TRIGA IPR-R1. The proposed design is based on a tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. Thus, the aim of this study is to verify the feasibility to establish the PGNAA method in IPR-R1 through theoretical study applying the Monte Carlo code. The feasibility of establishing the PGAA method at the IPR-R1 installations was evaluated through of the calculations of neutron flux, radioactive capture reaction rates and detection limits for some isotopes. According to the obtained results, it can be concluded that is possible to establish the PGAA method at the IPR-R1 reactor, even with some restrictions in its theoretical design calculated by MCNP. (author)

  11. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  12. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-07-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities.

  13. Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

    International Nuclear Information System (INIS)

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

    2015-01-01

    Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

  14. Measurements and calculations of reactivity for the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.; Maiorino, J.R.; Yamaguchi, M.

    1988-01-01

    This work shows a measurement of reactivity parameters, such as integral and diferential control rod worth, local void coefficient, and moderator temperature coefficient for the research reactor IEA-R1. The measured values were compared with those calculated through HAMMER-CITATION codes, having shown good agreement. (author) [pt

  15. Accident management strategies for VVER-1000 reactors. Part 1: text

    International Nuclear Information System (INIS)

    Sdouz, G.; Sonneck, G.; Pachole, M.

    1994-10-01

    This report describes the effect of different accident management strategies on the onset, development and end of the core-concrete-interaction as well as on the containment integrity for a TMLB'-type sequence in a Pressurized Water Reactor of the type VVER- 1000. Using the computer code MARCH3 the following strategies were investigated: (1) One or more Spray and LP ECC Systems available with and without coolers after 10 hours (2) Inclusion of the reactor pressure vessel testing facility room to the cavity (3) Containment venting (4) External water supply and (5) Different electric power restoration times. The results show that some of these accident management measures will maintain the containment integrity and reduce the source term drastically, others will reduce the source term rate. For some measures final conclusions can only be given after complete source term calculations have been performed. (authors)

  16. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  17. VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling

    International Nuclear Information System (INIS)

    Stewart, C.W.; Cuta, J.M.; Koontz, A.S.; Kelly, J.M.; Basehore, K.L.; George, T.L.; Rowe, D.S.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 1: Mathematical Modeling) explains the major thermal hydraulic models and supporting correlations in detail

  18. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  19. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  20. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  1. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  2. Anti-bacterial immunity to Listeria monocytogenes in allogeneic bone marrow chimera in mice

    International Nuclear Information System (INIS)

    Onoe, K.; Good, R.A.; Yamamoto, K.

    1986-01-01

    Protection and delayed-type hypersensitivity (DTH) to the facultative intracellular bacterium Listeria monocytogenes (L.m.) were studied in allogeneic and syngeneic bone marrow chimeras. Lethally irradiated AKR (H-2k) mice were successfully reconstituted with marrow cells from C57BL/10 (B10) (H-2b), B10 H-2-recombinant strains or syngeneic mice. Irradiated AKR mice reconstituted with marrow cells from H-2-compatible B10.BR mice, [BR----AKR], as well as syngeneic marrow cells, [AKR----AKR], showed a normal level of responsiveness to the challenge stimulation with the listeria antigens when DTH was evaluated by footpad reactions. These mice also showed vigorous activities in acquired resistance to the L.m. By contrast, chimeric mice that had total or partial histoincompatibility at the H-2 determinants between donor and recipient, [B10----AKR], [B10.AQR----AKR], [B10.A(4R)----AKR], or [B10.A(5R)----AKR], were almost completely unresponsive in DTH and antibacterial immunity. However, when [B10----AKR] H-2-incompatible chimeras had been immunized with killed L.m. before challenge with live L.m., these mice manifested considerable DTH and resistance to L.m. These observations suggest that compatibility at the entire MHC between donor and recipient is required for bone marrow chimeras to be able to manifest DTH and protection against L.m. after a short-term immunization schedule. However, this requirement is overcome by a preceding or more prolonged period of immunization with L.m. antigens. These antigens, together with marrow-derived antigen-presenting cells, can then stimulate and expand cell populations that are restricted to the MHC (H-2) products of the donor type

  3. Auroral kilometric radiation and magnetospheric substorm

    International Nuclear Information System (INIS)

    Morioka, Akira; Oya, Hiroshi

    1980-01-01

    The auroral kilometric radiation (AKR) and its relation to the development of the magnetospheric substorm have been studied based on the data obtained by JIKIKEN (EXOS-B) satellite. The occurrence of AKR is closely correlated to the intense UHR emission outside the plasmapause at the satellite position; the evidence clearly suggests that the development of the field aligned current system is associated with AKR generated at the upward current region and with the UHR emission at the downward current region. The drifting plasma due to the electric field that is generated in the magnetosphere at the moment of the magnetospheric substorm is derived from the frequency change of the plasma waves. The enhancement of the westward electric field in the duskside magnetosphere is detected simultaneously with the appearence of AKR. The altitude of the center of the AKR source region varies with intimate relation to the substorm activity suggesting that the generation of AKR is taking place in the region where the polar ionosphere and the magnetosphere are predominantly coupling through the precipitating or up going particles. From the fine structure of the dynamic spectra of AKR, it is suggested that the source of AKR might be closely related to the double layer type electric field along the magnetic field. (author)

  4. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter

    2013-01-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  5. Borated stainless steel storage project to the spent fuel of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio Carlos Iglesias; Madi Filho, Tufic; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks to the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating circumstances, the storage will have capacity for approximately six years. Whereas the estimated useful life of the IEA-R1 is around twenty years, it will be necessary to increase the storage capacity for the spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. This work objective is the project of high capacity storage for spent fuel elements, using borated stainless steel, to answer the Reactor IEA-R1 demand and the security requirements of the International Atomic Energy Agency. (author)

  6. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

    1985-02-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

  7. Cell surface response of chemically transformed, malignant mouse embryonal fibroblasts and human colon cancer cells to the maturation-promoting agent, N,N-dimethylformamide

    International Nuclear Information System (INIS)

    Marks, M.E.

    1985-01-01

    The lactoperoxidase/ 125 I radioiodination procedure was used to probe the cell surface of normal, nontransformed AKR-2B mouse embryo fibroblasts and malignant, permanently methylcholanthrene-transformed AKR-2B (AKR-MCA) cells to establish the relationship between cell surface changes and transformation/differentiation in this call system. AKR-MCA cells displayed surface alterations secondary to N,N-dimethylformamide (DFM)-promoted differentiation. Growth of AKR-MCA cells in DMF virtually eliminated the 85,000 and 63,000 molecular weight surface proteins susceptible to radioiodination and increased surface material of ∼200,000 molecular weight. Thus, surface profiles of DFM-treated AKR-MCA cells were essentially identical to those of nontransformed AKR-2B cells. Experimentation was extended to a cultured human colon cancer cell line (HCT MOSER). HCT MOSER cells exposed to DMF manifested marked, reversible morphological and surface changes which occurred as a function of time of growth in DMF and DMF concentration. Interestingly, material reactive with anti-fibronectin was found on the surfaces and in the culture medium of DFM-treated HCT MOSER cells

  8. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  9. Auroral kilometric radiation from transpolar arcs

    International Nuclear Information System (INIS)

    Pederson, B.M.; Pottelette, R.; Eliasson, L.; Murphree, J.S.; Elphinstone, R.D.; Bahnsen, A.; Jespersen, M.

    1992-01-01

    Observations from the Swedish satellite Viking allow the authors to study the relationship between auroral kilometric radiation (AKR) and discrete auroral features. Previous work has shown that AKR generation is most often associated with nightside aurora. They present wave data which show that under certain circumstances the source regions may also occur on discrete features, identified as transpolar arcs. The wave spectrograms detected during crossings or closest approaches to such sources exhibit structures similar to those observed during nightside AKR source crossings. Also, the associated ion beams and trapped conical electron populations with enhanced upward directed loss cones peak at comparable energies (∼1 keV)

  10. Space reactor electric systems: system integration studies, Phase 1 report

    International Nuclear Information System (INIS)

    Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

    1983-01-01

    This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

  11. Thermal performance of Egypt's research reactor core (ET-RR-1)

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.

    1986-01-01

    The steady state thermal performance of the ET-RR-1 core system is theoretically investigated by different models describing the heat flux and the coolant mass flow rate. The magnitude of the heat generated by a fuel element depends upon its position in the core. Normal and uniform distributions for heat flux and coolant mass flow rate are considered. The clad and coolant temperatures at different core positions are evaluated and compared with the experimental measurements at different operating conditions. The results indicated large discrepancy between the predicted and the experimental results. Therefore, the previous models and the experimental results are evaluated in order to develop the best model that describes the thermal performance of the ET-RR-1 core. The adapted model gives 99.5% significant confidence limit. The effect of increasing the heat flux or decreasing the mass flow rate by 20% from its maximum recommended operating condition is tested and discussed. Also, the thermal behaviour towards increasing the reactor power more than its maximum operating condition is discussed. The present work could also be used in extending the investigation to other PWR reactor operating conditions

  12. Calibration of new I and C at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, Martin; Jurickova, Monika

    2011-01-01

    The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

  13. Neutronics Design of Helical Type DEMO Reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Sagara, A.; Goto, T.; Yanagi, N.; Masuzaki, S.; Tamura, H.; Miyazawa, J.; Muroga, T., E-mail: teru@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

    2012-09-15

    Full text: Neutronics design study has been performed in a newly started conceptual design activity for a helical type DEMO reactor FFHR-d1. Features of the FFHR-d1 design are enlargement of the basic configurations of reactor components and extrapolation of plasma parameters from those of the helical type plasma experimental machine Large Helical Device (LHD) to achieve the highest feasibility. From the neutronics point of view, a blanket space of FFHR-d1 is severely limited at the inboard of the torus. This is due to the core plasma position shifting to the inboard side under the confinement condition extrapolated from LHD. The first step of the neutronics investigation using the MCNP code has been performed with a simple torus model simulating thin inboard blanket space. A Flibe+Be/Ferritic steel breeding blanket showed preferable performances for both tritium breeding and shielding, and has been adapted as a reference blanket system for FFHR-d1. The investigations indicate that a combination of a 15 cm thick breeding blanket, 55 cm thick WC+B4C shield, i.e., the blanket space of 70 cm, could suppress the fast neutron flux and nuclear heating in the helical coils to the design targets for the neutron wall loading of 1.5 MW/m{sup 2}. Since the outboard side can provide a large space for a 60 cm thick breeding blanket, a fully-covered tritium breeding ratio (TBR) of 1.31 has been obtained in the simple torus model. The neutronics design study has proceeded to the second step using a 3-D helical reactor model. The most important issue in the 3-D neutronics design is a compatibility with the helical divertor design. To achieve a higher TBR and shielding performance, the core plasma has to be covered by the breeding blanket layers as possible. However, the dimensions of the blanket layers are limited by magnetic field lines connecting an edge of the core plasma and divertor pumping ports. After repeating modification of the blanket configuration, the global TBR of 1

  14. HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.

    1987-05-01

    This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

  15. Generation of auroral kilometric radiation

    International Nuclear Information System (INIS)

    Green, J.L.

    1979-01-01

    Simultaneous observations between the Hawkeye spacecraft in the AKR emission cone and the low altitude polar orbiting spacecraft Triad and AE-D reveal that auroral kilometric radiation (AKR) is correlated with a variety of auroral particle precipitation in the evening to midnight local time sector. It is found that as the AKR intensity increases so does the integrated current sheet intensity of auroral zone field aligned currents observed by Triad of 257 simultaneous observations. Statistically, the linear correlation coefficient between the log of the AKR power flux and the log of the current sheet intensity is 0.57. Auroral kilometric radiation observations from Hawkeye during low altitude (2.0 to 2.5 R/sub E/) auroral zone passes reveal that intense AKR has a low frequency cutoff near the local electron gyrofrequency (f/sub g/ - ) with maximum electric field strengths as large as 12 mV/m. The large electric fields observed near f/sub g/ - are consistent with high altitude observations of AKR using a simple 1/R 2 scaling indicating that the kilometric radiation in or near the average source region is almost completely electromagnetic. The results presented in this study indicate that kilometric radiation is generated by inverted-V electron distribution functions in a direct coupling mechanism between particle energy and R-X mode electromagnetic waves in the region of the auroral zone where f/sub g/ - >> f/sub p/ -

  16. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

    Described is what occurred in the reactors of Fukushima Daiichi Nuclear Power Plant at the Tohoku earthquake and tsunami (Mar. 11, 2011) from the aspect of engineering science. The tsunami attacked the Plant 1 hr after the quake. The Plant had reactors in buildings no.1-4 at 10 m height from the normal sea level which was flooded by 1.5-5.5 m high wave. All reactors in no.1-6 in the Plant were the boiling water type, and their core nuclear reactions were stopped within 3 sec due to the first quake by control rods inserted automatically. Reactors in no.1-5 lost their external AC power sources by the breakdown and subsequent submergence (no.1-4) of various equipments and in no.1, 2 and 4, the secondary DC power was then lost by the battery death. Although the isolation condenser started to cool the reactor in no.1 after DC cut, its valve was then kept closed to heat up the reactor, leading to the reaction of heated Zr in the fuel tube and water to yield H 2 which was accumulated in the building: the cause of hydrogen explosion on 12th. The reactor in no.2 had the reactor core isolation cooling system (RCIC) which operated normally for few hrs, then probably stopped to heat up the reactor, resulting in meltdown of the core but no explosion occurred because of the opened door of the blowout panel on the wall by the blast of no.1 explosion. The reactor in no.3 had RCIC and high pressure coolant injection system, but their works stopped to result in the core damage and H 2 accumulation leading to the explosion on 14th. The reactor in no.4 had not been operated because of its periodical annual examination, but was explored on 15th, of which cause was thought to be due to backward flow of H 2 from no.3. Finally, the author discusses about this accident from the industrial aspect of the design of safety level (defense in depth) on international views, and problems and tasks given. (T.T.)

  17. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  18. The role of SASSYS-1 in LMR [Liquid Metal Reactor] safety analysis

    International Nuclear Information System (INIS)

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs

  19. Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

    2009-01-01

    Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

  20. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  1. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  2. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  3. Multipurpose RTOF Fourier diffractometer at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Maayouf, R.M.A.; Tiitta, A.T.

    1993-09-01

    The present work represents a further study of the basic RTOF Fourier multipurpose diffractometer, to start with, at the ET-RR-1 reactor. The functions of the suggested arrangement are thoroughly discussed and the possibilities if its expansion are also assessed. The flexibility of the arrangement allows its further expansion both for stress measurement at 90 deg. scattering angle with two detector banks at opposite sides of the incident beam and for operation in the transmission diffraction mode. (orig.). (19 refs., 10 figs., 1 tab.)

  4. Competitive proliferation in the hematopoietic tissues of irradiated hybrid mice engrafted with parental bone marrow and spleen

    International Nuclear Information System (INIS)

    Muramatsu, S.; Monnot, P.; Duplan, J.F.

    1976-01-01

    e kinetics of growth and differentiation of hematopoietic stem cells differ markedly according to their origin. A study of the ability of CFU from bone marrow (BM) or spleen to repopulate hemopoietic organs has been carried out in lethally irradiated mice restored with BM cells admixed with spleen cells bearing different chromosomal markers. Hemopoietic cells originating from AKR (40 acbrocentrics) and AKR/T1ALD (36 acrocentrics + 2 metacentrics) mice were engrafted into lethally irradiated (AKR x AKR/T1ALD)F1 or (C3H x AKR/T1ALD)F1 hybrid recipients. Within 10 days, the BM-derived elements outnumbered the spleen-derived population in BM and spleen. This held even when the number of injected spleen-CFU was twice that of BM-CFU. This difference of growth rate subsided within 20 days. The first cells to reappear in the thymus bore the recipient karyotype (endoregeneration); they were later replaced by BM-derived elements but spleen-derived cecells were never present in thymus in the case of competitive engraftment. In contrast, the lymph node cells bore the BM karyotype as well as the spleen karyotype. Injecting the spleen cells 3 days prior to the BM cells partially counterbalanced the overgrowth of the BM-derived elements in the BM and spleen but did not affect the thymic repopulation which remained strictly derived from BM-CFU. When mice were injected only with BM-CFU, or only with spleen-CFU, BM-derived cells were found in the thymus as early as 10-12 days after engraftment, whereas the spleen-derived cells did not appear in the thymus until days 18-20. (author)

  5. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  6. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  7. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    International Nuclear Information System (INIS)

    Auterinen, Iiro; Salmenhaara, Seppo

    2008-01-01

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  8. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  9. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  10. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  11. Maintenance of reactor recirculation pumps [Paper No.: II-1

    International Nuclear Information System (INIS)

    Ansari, M.A.; Bhat, K.P.

    1981-01-01

    At Tarapur Atomic Power Station (TAPS), two reactor recirculation pumps are provided, one each for the two reactor units. The performance of pumps has been uniformly good; however, leakage through the cartridge type, two stage, mechanical seals which are installed on these pumps was encountered on few occasions. The paper describes the leakage problems, identification of certain design deficiencies and rectification carried out at TAPS for overcoming these problems. (author)

  12. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  13. Design and development of weld inspection manipulator for reactor pressure vessel of TAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, H.; Singh, J.P.; Ranjon, R.; Kulkarni, M.P.; Patel, R.J.

    2013-01-01

    The reactor pressure vessel (RPV) of TAPS-1 BWR contains six longitudinal and four circumferential welds. Periodical in-service inspection of these weld joints has been a regulatory issue pending for long. In the 22 nd refuelling outage in July 2012 the inspection of L1-1, L1-2 longitudinal welds as well as their junctions with C1 circumferential weld were proposed to be done using ultrasonic technique. Approaching these welds from OD side of the RPV is a difficult and tedious task. Therefore it was decided to examine these welds from ID side of the RPV by filling the cavity with water and approaching the RPV from top. No technology was locally available to take the probes at a depth of 10-12 m under water. NPCIL approached RTD, BARC to develop an underwater manipulator to accomplish this task. RTD took up this work as a challenge and came out with the design of manipulator. The weld inspection manipulator (WIM) was fabricated on a war foot basis, tested and successfully implemented in the reactor for the first time in TAPS history. The entire activity was completed in three months time. This article gives the details of design, manufacturing, performance testing, qualification trials and implementation of WIM in the reactor. Ultrasonic testing techniques were developed by QAD, BARC which are not covered in this article. (author)

  14. Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

    2005-01-01

    A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

  15. Neutronics analysis of Nigerian Research Reactor-1

    International Nuclear Information System (INIS)

    Azande, T.S.; Balogun, G.I.

    2010-01-01

    Feasibility studies for the conversion of the Nigerian Research Reactor-1 (NIRR-1) have been performed using WIMS and CITATION codes (Azande et al, 2009 and Balogun, 2003) at the Centre for Energy Research and Training (CERT), Ahmadu Bello University, Zaria Kaduna State. In this work, the neutronics analysis of NIRR-1 core concerning mass loading of U-235 in the core, shut down margin (SDM), safety reactivity factor (SRF), control rod worth, and control rod critical depth of insertion were investigated at low enrichment. Two fuel types (UAl 4 and UO 2 ) were considered and the uranium densities required for the conversion of NIRR-1 core to low enrichment were computed to be 1201g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1274 g/cc with 15% enrichment, 1448 g/cc with 10% enrichment for UAl 4 fuel type and 1141g/cc with 20% enrichment, 1144 g/cc with 19.75% enrichment, 1216 g/cc with 15% enrichment, and 1389 g/cc with 10% enrichment for UO 2 fuel type. Signi ficantly, higher uranium densities are required to convert NIRR-1 from HEU to LEU - indicating a drastic review of the NIRR-1 core.

  16. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  17. Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

    International Nuclear Information System (INIS)

    Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

    2014-01-01

    Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

  18. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO 2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPR TM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1 TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENA TM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENA TM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  19. Isolation and primary structural analysis of two conjugated polyketone reductases from Candida parapsilosis.

    Science.gov (United States)

    Hidalgo, A R; Akond, M A; Kita, K; Kataoka, M; Shimizu, S

    2001-12-01

    Two conjugated polyketone reductases (CPRs) were isolated from Candida parapsilosis IFO 0708. The primary structures of CPRs (C1 and C2) were analyzed by amino acid sequencing. The amino acid sequences of both enzymes had high similarity to those of several proteins of the aldo-keto-reductase (AKR) superfamily. However, several amino acid residues in the putative active sites of AKRs were not conserved in CPRs-C1 and -C2.

  20. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  1. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  2. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  3. Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

    <