Operating experience from Swedish nuclear power plants 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

The total production of electricity from Swedish nuclear power plants was 65.6 TWh during 2002, which is a decrease compared to 2001. The energy capability factor for the 11 Swedish reactors averaged 80.8%. The PWRs at Ringhals averaged 87.6%, while the BWRs, not counting Oskarshamn 1, reached 89.2%. No events, which in accordance to conventions should be reported to IAEA, have occurred during 2002. Operational statistics are presented for each Swedish reactor. The hydroelectric power was 66 TWh, 16% lower than 2000. Wind power contributed 0.5 TWh, and remaining production sources, mainly from solid fuel plants combined with district heating, contributed 10.9 TWh. The electricity generation totalled 143 TWh, considerably less than the record high 2001 figure of 158.7 TWh. The preliminary figures for export were 14.8 TWh and and for import 20.1 TWh.

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

Energy Technology Data Exchange (ETDEWEB)

Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

2007-08-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

Fundamental design bases for independent core cooling in Swedish nuclear power reactors

International Nuclear Information System (INIS)

Jelinek, Tomas

2015-01-01

New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

The r-Java 2.0 code: nuclear physics

Science.gov (United States)

Kostka, M.; Koning, N.; Shand, Z.; Ouyed, R.; Jaikumar, P.

2014-08-01

Aims: We present r-Java 2.0, a nucleosynthesis code for open use that performs r-process calculations, along with a suite of other analysis tools. Methods: Equipped with a straightforward graphical user interface, r-Java 2.0 is capable of simulating nuclear statistical equilibrium (NSE), calculating r-process abundances for a wide range of input parameters and astrophysical environments, computing the mass fragmentation from neutron-induced fission and studying individual nucleosynthesis processes. Results: In this paper we discuss enhancements to this version of r-Java, especially the ability to solve the full reaction network. The sophisticated fission methodology incorporated in r-Java 2.0 that includes three fission channels (beta-delayed, neutron-induced, and spontaneous fission), along with computation of the mass fragmentation, is compared to the upper limit on mass fission approximation. The effects of including beta-delayed neutron emission on r-process yield is studied. The role of Coulomb interactions in NSE abundances is shown to be significant, supporting previous findings. A comparative analysis was undertaken during the development of r-Java 2.0 whereby we reproduced the results found in the literature from three other r-process codes. This code is capable of simulating the physical environment of the high-entropy wind around a proto-neutron star, the ejecta from a neutron star merger, or the relativistic ejecta from a quark nova. Likewise the users of r-Java 2.0 are given the freedom to define a custom environment. This software provides a platform for comparing proposed r-process sites.

Operating experience from Swedish nuclear power plants 2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

The total production of electricity from Swedish nuclear power plants was 69.2 TWh during 2001, which is an increase of more than 25% compared to 2000. The hydroelectric power production increased to 78.3 TWh, 22% more than during a normal year, i.e. a year with average rainfall. Wind power contributed 0.5 TWh, and remaining production sources, mainly from solid fuel plants combined with district heating, contributed 9.6 TWh. The electricity generation totalled 157.6 TWh, the highest annual production to date. The preliminary figures for export were 18.5 TWh and and for import 11.1 TWh. Operational statistics are presented for each Swedish reactor. Two events, given INES level 1 rating, are reported from Barsebaeck 2 and Ringhals 2.

Operating experience from Swedish nuclear power plants 2001

International Nuclear Information System (INIS)

2002-01-01

The total production of electricity from Swedish nuclear power plants was 69.2 TWh during 2001, which is an increase of more than 25% compared to 2000. The hydroelectric power production increased to 78.3 TWh, 22% more than during a normal year, i.e. a year with average rainfall. Wind power contributed 0.5 TWh, and remaining production sources, mainly from solid fuel plants combined with district heating, contributed 9.6 TWh. The electricity generation totalled 157.6 TWh, the highest annual production to date. The preliminary figures for export were 18.5 TWh and and for import 11.1 TWh. Operational statistics are presented for each Swedish reactor. Two events, given INES level 1 rating, are reported from Barsebaeck 2 and Ringhals 2

Restart of R reactor at SRP

International Nuclear Information System (INIS)

McDonell, W.R.

1983-01-01

Restart of the Savannah River R-Reactor is an alternative to L-Reactor operation for increased production of defense nuclear material. R-Reactor was shut down in 1964 after 11-years operation and has been on standby for 19 years. This report presents a description of R-Reactor operation to serve as a basis for analysis of environmental impacts after restoration to meet current SRP performance standards. R-Reactor operation would differ from L-Reactor operation principally in discharge and recycle of effluent cooling water to Par Pond, rather than direct discharge to the Savannah River by way of Steel Creek. Significant differences in environmental effects could result. A costly renovation program would be required to restore R-Reactor to operability, and the reactor could not contribute to material production before about 1989

Word-initial /r/-clusters in Swedish speaking children with typical versus protracted phonological development.

Science.gov (United States)

Lundeborg Hammarström, Inger

2018-01-01

The present study investigated word-initial (WI) /r/-clusters in Central Swedish-speaking children with and without protracted phonological development (PPD). Data for WI singleton /r/ and singleton and cluster /l/ served as comparisons. Participants were twelve 4-year-olds with PPD and twelve age- and gender-matched typically developing (TD) controls. Native speakers audio-recorded and transcribed 109 target single words using a Swedish phonology test with 12 WI C+/r/-clusters and three WI CC+/r/-clusters. The results showed significantly higher match scores for the TD children, a lower match proportion for the /r/ targets and for singletons compared with clusters, and differences in mismatch patterns between the groups. There were no matches for /r/-cluster targets in the PPD group, with all children except two in that group showing deletions for both /r/-cluster types. The differences in mismatch proportions and types between the PPD group and controls suggests new directions for future clinical practice.

On flux effects in a low alloy steel from a Swedish reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

2017-02-15

This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

Development, implementation, and experiences of the Swedish spent fuel and waste sea transportation system

International Nuclear Information System (INIS)

Gustafsson, B.; Dybeck, P.; Pettersson, S.

1989-01-01

In Sweden, electrical production from the first commercial nuclear plant commenced in 1972, i.e. 17 years ago. There are now 12 nuclear reactors in operation, the last two were connected to the grid in fall 1985. These 12 reactors produced about 50% of the present electrical demand in Sweden. The remaining 50% are mainly covered by hydro power stations. The operating record for the Swedish reactors has generally been very good. Nevertheles, the Swedish parliament has taken a decision, that nuclear power shall be phased out from the Swedish system not later than the year 2010. Many of them - to use a mild expression-question the wisdom of this decision. The efforts in the waste management area will, however, be given a continued high priority. The primary responsibility for the management of nuclear waste lies with the waste producer. In order to achieve a good coordination and an effective management the four Swedish nuclear power utilities have delegated these responsibilities to the jointly owned Swedish Nuclear Fuel and Waste Management Co., SKB. This means that SKB is responsible for measures required for the implementation of the national nuclear waste management program such as planning, design, construction and operation of waste facilities including the necessary R and D work. The responsibility of the nuclear power utilities also includes the financing of the waste management program. A special funding system, controlled by the authorities, has been established for this purpose

Experience in plant transients. The Swedish RKS program

International Nuclear Information System (INIS)

Bento, J.P.

1983-09-01

A data-base for reactor operation experience is presented. The input comes from utilities in 14 countries. From experience with the Swedish reactors, trends have been extracted. Using the number of operational scrams as a measure of reactor management, there seem to be a maximum at early reactor life, followed by a decreasing trend after 2 years. This seems to be true for all reactors in the programme. There is even a decrease in the number of scrams with further reactor generations. Causes for events and for scrams are evaluated. (Aa)

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

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)

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

Variation in the MC4R gene is associated with bone phenotypes in elderly Swedish women.

Directory of Open Access Journals (Sweden)

Gaurav Garg

Full Text Available Osteoporosis is characterized by reduced bone mineral density (BMD and increased fracture risk. Fat mass is a determinant of bone strength and both phenotypes have a strong genetic component. In this study, we examined the association between obesity associated polymorphisms (SNPs with body composition, BMD, Ultrasound (QUS, fracture and biomarkers (Homocysteine (Hcy, folate, Vitamin D and Vitamin B12 for obesity and osteoporosis. Five common variants: rs17782313 and rs1770633 (melanocortin 4 receptor (MC4R; rs7566605 (insulin induced gene 2 (INSIG2; rs9939609 and rs1121980 (fat mass and obesity associated (FTO were genotyped in 2 cohorts of Swedish women: PEAK-25 (age 25, n = 1061 and OPRA (age 75, n = 1044. Body mass index (BMI, total body fat and lean mass were strongly positively correlated with QUS and BMD in both cohorts (r(2 = 0.2-0.6. MC4R rs17782313 was associated with QUS in the OPRA cohort and individuals with the minor C-allele had higher values compared to T-allele homozygotes (TT vs. CT vs.100 vs. 103 vs. 103; p = 0.002; (SOS: 1521 vs. 1526 vs. 1524; p = 0.008; (Stiffness index: 69 vs. 73 vs. 74; p = 0.0006 after adjustment for confounders. They also had low folate (18 vs. 17 vs. 16; p = 0.03 and vitamin D (93 vs. 91 vs. 90; p = 0.03 and high Hcy levels (13.7 vs 14.4 vs. 14.5; p = 0.06. Fracture incidence was lower among women with the C-allele, (52% vs. 58%; p = 0.067. Variation in MC4R was not associated with BMD or body composition in either OPRA or PEAK-25. SNPs close to FTO and INSIG2 were not associated with any bone phenotypes in either cohort and FTO SNPs were only associated with body composition in PEAK-25 (p≤0.001. Our results suggest that genetic variation close to MC4R is associated with quantitative ultrasound and risk of fracture.

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

The use of plutonium in Swedish reactors

International Nuclear Information System (INIS)

Forsstroem, H.

1982-09-01

The report deals with the utilization of plutonium in Swedish nuclear power plants. The plutonium content of the mixed oxide fuel will normally be 3-7 per cent. The processing of spent nuclear fuel will produce about 6 ton plutonium. The use of mixed oxide fuel in Forsmark 3 and Oskarshamn 3 is discussed. The fuel cycle will start with the manufacturing of the fuel elements abroad and proceeds with transport and utilization, storing of spent fuel about 40 years in Sweden followed by direct disposal. The manufacture and use of mixed oxide (MOX) fuel is based on well-known techniques. Approximately 20 000 MOX fuel rods have been irradiated and the fuel is essentially equivalent to uranium oxide fuel. 30-50 per cent of the core may be composed of MOX-fuel without any effect on the operation and safety of the reactor which has been originally designed for uranium fuel. The evaluation of international fuel cycle (INFCE) states that the proliferation risks are very small. The recycling of plutonium will reduce demand for enriched uranium and the calculations show that 6.3 ton plutonium will replace the enrichment of 600 ton natural uranium. (G.B.)

Operating experience from Swedish nuclear power plants 2003

International Nuclear Information System (INIS)

2004-01-01

In safety terms, operations of the Swedish nuclear power plants in 2003 can be summarized as having ben good, with a few exceptions: The thermal mixer problem at Barsebaeck-2; The Highest Permissible Limit Value excursion at OKG-3, which subjected the reactor pressure vessel to a too rapid temperature change; and An INES class 1 incident at Ringhals-1. The Barsebaeck and Ringhals events were not of such seriousness as to represent a threat to reactor safety, but they both had the effect of causing the Nuclear Power Inspectorate to question safety cultures at the plants. The mixer event resulted in a considerable production loss, with the reactor being shut down twice, making a total of five months. OKG-3 was shut down for almost two months during the autumn. Despite the above, production from Swedish NPPs was much the same as during 2002. Total electricity production amounted to 65 TWh (65.2 TWh in 2002). On the average the energy availability of the eleven reactors was 79%. The PWRs at Ringhals had an average energy availability of 89%, while the BWRs reached 76%

Microstructure in Zircaloy Creep Tested in the R2 Reactor

International Nuclear Information System (INIS)

Pettersson, Kjell

2004-12-01

Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

Operational Experience from Swedish nuclear power plants 1996

International Nuclear Information System (INIS)

1997-01-01

A summary of two pages is given for each Swedish reactor with data on availability, scrams, radiation doses and important events during 1996. Special reports are presented on the following issues: Reactor core spray system inoperable at OKG-2, Containment pressure relief system incorrectly closed at Forsmark-1, Isolation condenser blocked for residual heat and continued operation with defective isolation valve at OKG-1; and Degraded pressure suppression function of the containment at Barsebaeck-2

The impact of the Swedish massage on the kinesthetic differentiation in healthy individuals.

Science.gov (United States)

Mustafa, Kamil; Furmanek, Mariusz Pawel; Knapik, Aleksandra; Bacik, Bogdan; Juras, Grzegorz

2015-03-01

Swedish massage is one of the common treatments to provide optimal start and readiness of athletes. The ability of kinesthetic differentiation (KD) is crucial in sport performance. This skill allows to adapt demanded muscle forces to optimize the motor tasks, and it is responsible for the precision. In the literature, there is no evidence how Swedish massage influences the kinesthetic differentiation. The objective of the study was to evaluate the impact of Swedish massage on the kinesthetic differentiation and muscle strength of hand grip. Thirty participants took part in this investigation (17 women and 13 men). The assessment consisted of KD tests conducted on the dominant (DH) and nondominant hand (NDH) after 15 minutes of hand and forearm Swedish massage. The procedure consisted of 13 trials for each extremity. The first three were done for 100% of the participants' capabilities (Fmax), the next five trials were done using 50% of maximum force (50% of Fmax), and in the last five trials, the participants tried to use only 50% of their previous force (1/2 of 50%). Finally, the absolute force production error (FPE) was calculated for 50% (FPE_50%) and 25% (FPE_25%). The two-way repeated measure analysis of variance ANOVA did not reveal any statistically significant changes in maximal strength grip and KD between pre- and postmassage intervention in both DH and NDH hand. Correlations showed strong relationship between pre- and postmassage for maximum force (r = 0.92, p = .01 for DH, and r = 0.94, p = .01 for NDH), and only for the FPE_50% (r = 0.67, p = .01 for DH, and r = 0.71, p = .01 for NDH). The results obtained indicated that the application of the Swedish massage did not affect the kinesthetic differentiation in this particular young adult group.

The Impact of the Swedish Massage on the Kinesthetic Differentiation in Healthy Individuals

Science.gov (United States)

Mustafa, Kamil; Furmanek, Mariusz Pawel; Knapik, Aleksandra; Bacik, Bogdan; Juras, Grzegorz

2015-01-01

Background: Swedish massage is one of the common treatments to provide optimal start and readiness of athletes. The ability of kinesthetic differentiation (KD) is crucial in sport performance. This skill allows to adapt demanded muscle forces to optimize the motor tasks, and it is responsible for the precision. In the literature, there is no evidence how Swedish massage influences the kinesthetic differentiation. Purpose: The objective of the study was to evaluate the impact of Swedish massage on the kinesthetic differentiation and muscle strength of hand grip. Methods: Thirty participants took part in this investigation (17 women and 13 men). The assessment consisted of KD tests conducted on the dominant (DH) and nondominant hand (NDH) after 15 minutes of hand and forearm Swedish massage. The procedure consisted of 13 trials for each extremity. The first three were done for 100% of the participants’ capabilities (Fmax), the next five trials were done using 50% of maximum force (50% of Fmax), and in the last five trials, the participants tried to use only 50% of their previous force (1/2 of 50%). Finally, the absolute force production error (FPE) was calculated for 50% (FPE_50%) and 25% (FPE_25%). Results: The two-way repeated measure analysis of variance ANOVA did not reveal any statistically significant changes in maximal strength grip and KD between pre- and postmassage intervention in both DH and NDH hand. Correlations showed strong relationship between pre- and postmassage for maximum force (r = 0.92, p = .01 for DH, and r = 0.94, p = .01 for NDH), and only for the FPE_50% (r = 0.67, p = .01 for DH, and r = 0.71, p = .01 for NDH). Conclusions: The results obtained indicated that the application of the Swedish massage did not affect the kinesthetic differentiation in this particular young adult group. PMID:25780470

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

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)

Technology and costs for decommissioning of Swedish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-06-01

The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs.

Technology and costs for decommissioning of Swedish nuclear power plants

International Nuclear Information System (INIS)

1994-06-01

The decommissioning study for the Swedish nuclear power plants has been carried out during 1992 to 1994 and the work has been led by a steering group consisting of people from the nuclear utilities and SKB. The study has been focused on two reference plants, Oskarshamn 3 and Ringhals 2. Oskarshamn 3 is a boiling water reactor (BWR) and Ringhals 2 is a pressurized water reactor (PWR). Subsequently, the result from these plants have been translated to the other Swedish plants. The study gives an account of the procedures, costs, waste quantities and occupational doses associated with decommissioning of the Swedish nuclear power plants. Dismantling is assumed to start immediately after removal of the spent fuel. No attempts at optimization, in terms of technology or costs, have been made. The nuclear power plant site is restored after decommissioning so that it can be released for use without restriction for other industrial activities. The study shows that a reactor can be dismantled in about five years, with an average labour force of about 150 persons. The maximum labour force required for Oskarshamn 3 has been estimated to about 300 persons. This peak load occurred the first years but is reduced to about 50 persons during the demolishing of the buildings. The cost of decommissioning Oskarshamn 3 has been estimated to be about MSEK 940 in January 1994 prices. The decommissioning of Ringhals 2 has been estimated to be MSEK 640. The costs for the other Swedish nuclear power plants lie in the range MSEK 590-960. 17 refs, 21 figs, 15 tabs

Operating experience from Swedish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-06-01

The total production of electricity from Swedish nuclear power plants was 70.5 TWh during 1998, which is the second highest yearly production ever. Production losses due to low demand totaled 5.1 TWh combined for all twelve units and production losses due to coastdown operation totaled an additional 0.5 TWh. The reason for this low power demand was a very good supply of water to the hydropower system. Hydroelectric power production was 73.6 TWh, an increase by roughly 5 TWh since 1997. Hence, the hydroelectric power production substantially exceeded the 64 TWh expected during a normal year, i.e. a year with average rainfall. Remaining production sources, mainly fossil fuel electricity production combined with district heating, contributed with 10 TWh. The total electricity production was 154.2 TWh, the highest yearly production ever. The total electricity consumption including transmission losses was 143.5 TWh. This is also the highest consumption ever and an increase by one percent compared to 1997. The preliminary net result of the electric power trade shows a net export by 10.7 TWh. The figures above are calculated from the preliminary production results. A comprehensive report on electric power supply and consumption in Sweden is given in the 1998 Annual Report from the Swedish Power Association. Besides Oskarshamn 1, all plants have periodically been operated in load-following mode, mostly because of the abundant supply of hydropower. The energy availability for the three boiling water reactors at Forsmark averaged 93.3 % and for the three pressure water reactors at Ringhals 91.0 %, both figures are the highest ever noted. In the section `Special Reports` three events of importance to safety that occurred during 1998 are reported. The events were all rated as level 1 according to the International Nuclear Event Scale (INES) Figs, tabs.; Also available in Swedish

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

(1R,2R,3R,4R,5S-2,3-Bis[(2S′-2-acetoxy-2-phenylacetoxy]-4-azido-1-[(2,4-dinitrophenylhydrazonomethyl]bicyclo[3.1.0]hexane

Directory of Open Access Journals (Sweden)

Robert McDonald

2008-02-01

Full Text Available In the title compound, C38H29N7O12, the five-membered ring adopts an envelope conformation in which the `flap' is cis to the cyclopropane group. This conformation is similar to those of other bicyclo[3.1.0]hexane analogues for which crystal structures have been reported. The absolute configuration of the stereogenic centers on the cyclopentane ring, as determined by comparison with the known configurations of the stereogenic centers in the (2S-2-acetoxy-2-phenylacetoxy groups, is 1(R, 2(R, 3(R, 4(R and 5(S. An intramolecular N—H...O hydrogen bond is present.

Sources of gamma radiation in a reactor core

Energy Technology Data Exchange (ETDEWEB)

Roos, Matts

1959-05-15

In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

International Nuclear Information System (INIS)

Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

1969-01-01

1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

Energy Technology Data Exchange (ETDEWEB)

Rumis, D; Leszczynski, F

1991-12-31

An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

Focal F0 peak shape and sentence mode in Swedish

DEFF Research Database (Denmark)

Ambrazaitis, Gilbert; Buanzur, Tuarik C.; Niebuhr, Oliver

2015-01-01

Shape characteristics of rising-falling accentual F0 peaks of Stockholm Swedish Accent I words in narrow focus are studied in a corpus of 287 read sentences. The corpus includes statements and three types of polar questions. Results reveal a clear effect of sentence mode on the shape of the accen...

Electrochemical aspects on corrosion in Swedish reactor containments; Elektrokemiska aspekter paa korrosion i svenska reaktorinneslutningar

Energy Technology Data Exchange (ETDEWEB)

Ullberg, Mats [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

2006-10-15

Post-stressed concrete is used in all Swedish nuclear reactor containments. Steel in concrete is normally protected from corrosion by the highly alkaline pore solution in concrete. A passive film develops on the surface of steel in contact with the pore solution. However, corrosion may still occur under special circumstances. It is therefore desirable to monitor the corrosion status of the containment. A review of the corrosion experience with steel in concrete strongly suggests that the potential problem of most concern for the Swedish reactor containments is cavity formation during grouting of tendons and of penetrations in the containment wall. Cavities break the contact between alkaline grout and steel. Corrosion is then possible, provided the relative humidity is high enough. Normal methods for inspection of the corrosion status of steel reinforcement in concrete are not applicable to very heavy structures like reactor containments. Since inspections are difficult to carry out, it is important that they be focused on the most susceptible portions of the containment. This report is an attempt to assemble potentially useful background information. The original intention was to focus on electrochemical methods of investigation. When it was realized that the potential use of electrochemical methods was limited, the scope of the review was broadened. The present as well as previous investigations indicate that nondestructive testing of grouted tendons is the outstanding problem in the condition assessment of Swedish nuclear reactor containments. Grouted tendons are also used in a very large number of bridges built since the early 1950s. The experience gained in connection with bridges has therefore been investigated. The need for a testing method for grouted tendons in bridges has long been strongly felt and development work has been in progress since the early 1970-ies, for example within the Strategic Highway Research Project in the Unite States. Potential

Safety Assessment - Swedish Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

1996-12-31

After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

Safety Assessment - Swedish Nuclear Power Plants

International Nuclear Information System (INIS)

Kjellstroem, B.

1996-01-01

After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs

The nuclear question at the start of the '80s: the breeder reactor

International Nuclear Information System (INIS)

Owen, R.; Svensson, B.

1980-01-01

The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO 2 -UO 2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

Swedish Government Minister at CERN

CERN Document Server

2008-01-01

The Swedish Minister for Higher Education and Research recently visited CERN. The Swedish Minister was greeted by Swedish scientists working at CERN. Signing of the Swedish Computing Memorandum of Understanding. Pär Omling, Director-General of the Swedish Research Council (left), and Jos Engelen, CERN’s Chief Scientific Officer. Lars Leijonborg, the Swedish Minister for Higher Education and Research, was welcomed to CERN by Director-General Robert Aymar on 10 March. After an introduction to the Laboratory’s activities, the Minister was given guided tours of the control room, the ATLAS surface hall and experiment cavern and the adjoining LHC tunnel. Mr Leijonborg was then greeted by Swedish scientists and given an overview of the Swedish research programme at CERN. Five Swedish university groups are taking part in LHC research. Swedish universities are notably involved in the manufacture of parts for the sub-detectors of AT...

Enterprise 2.0 - hvordan får du succes?

DEFF Research Database (Denmark)

Ivang, Reimer

2009-01-01

Det seneste år til halvandet har der været konstant stigende fokus og interesse for begrebet Enterprise 2.0. En interesse som kan ses af den konstant stigende litteratur som beskæftiger sig med emnet. Men er det noget for dig, for din organisation? Få svaret her.......Det seneste år til halvandet har der været konstant stigende fokus og interesse for begrebet Enterprise 2.0. En interesse som kan ses af den konstant stigende litteratur som beskæftiger sig med emnet. Men er det noget for dig, for din organisation? Få svaret her....

Education for the nuclear power industry: Swedish perspective

International Nuclear Information System (INIS)

Blomgren, J.

2005-01-01

In the Swedish nuclear power industry staff, very few newly employed have a deep education in reactor technology. To remedy this, a joint education company, Nuclear Training and Safety Center (KSU), has been formed. To ensure that nuclear competence will be available also in a long-term perspective, the Swedish nuclear power industry and the Swedish Nuclear Power Inspectorate (SKI) have formed a joint center for support of universities, the Swedish Nuclear Technology Center (SKC). The activities of these organisations, their links to universities, and their impact on the competence development for the nuclear power industry will be outlined. (author)

Experience with the RE fuel transition at the Studsvik R2 reactor

International Nuclear Information System (INIS)

Pazsit, I.; Saltvedt, K.

1991-01-01

Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

Changes in control room at Swedish nuclear power plants

International Nuclear Information System (INIS)

Kecklund, Lena

2005-09-01

The Swedish nuclear power plants were commissioned during a period between 1972 and 1985 and the instrumentation and control equipment are basically from that period. For several years there have been plans made for changes in all the nuclear power plants and to a certain extent the changes in control equipment and monitoring rooms have also been implemented. The object of this project was to make a comprehensive review of the changes in control room design implemented in the Swedish nuclear power plants and to describe how the MTO- (Man-Technology-Organisation) and (Man-Machine-Interface) -issues have been integrated in the process. The survey is intended to give an overall picture of the changes in control room design and man-machine-interface made in the Swedish control rooms, in order to get a deeper knowledge of the change management process and its results as well as of the management of MTO-issues in these projects. The units included in this survey are: Oskarhamn reactor 2 and 3; Ringhals reactor 2, 3 and 4; Forsmark reactor 1, 2 and 3. The Oskarshamn 1 unit has not been included in this report as it has recently undergone an extensive modernisation program as well as a detailed inspection by the SKI (Swedish Nuclear Power Inspectorate). At Ringhals 2 the modernisation work is carried out at present and the unit is also subjected to extensive inspection activities carried out by SKI and is therefore not part of this survey. This report also includes a short description of relevant standards and requirements. Then follows a presentation of the results of the plant survey, presented as case studies for three companies OKG, Ringhals and FKA. Control room changes are summarized as well as the results on specific MTO issues which has been surveyed. In all the power companies there is a joint way of working with projects concerning plant modifications. This process is described for each company separately. In the concluding of the report the strengths and

Summary of IEA-R1 research a reactor licensing related to its power increase from 2 to 10 MW

International Nuclear Information System (INIS)

1989-04-01

This work is a summary of IEA-R1 research reactor licensing related to its power increase from 2 to 10 MW. It reports also safety requirements, fuel elements, and reactor control modifications inherent to power increase. (A.C.A.S.)

Swedish nuclear waste efforts

International Nuclear Information System (INIS)

Rydberg, J.

1981-09-01

After the introduction of a law prohibiting the start-up of any new nuclear power plant until the utility had shown that the waste produced by the plant could be taken care of in an absolutely safe way, the Swedish nuclear utilities in December 1976 embarked on the Nuclear Fuel Safety Project, which in November 1977 presented a first report, Handling of Spent Nuclear Fuel and Final Storage of Vitrified Waste (KBS-I), and in November 1978 a second report, Handling and Final Storage of Unreprocessed Spent Nuclear Fuel (KBS II). These summary reports were supported by 120 technical reports prepared by 450 experts. The project engaged 70 private and governmental institutions at a total cost of US $15 million. The KBS-I and KBS-II reports are summarized in this document, as are also continued waste research efforts carried out by KBS, SKBF, PRAV, ASEA and other Swedish organizations. The KBS reports describe all steps (except reprocessing) in handling chain from removal from a reactor of spent fuel elements until their radioactive waste products are finally disposed of, in canisters, in an underground granite depository. The KBS concept relies on engineered multibarrier systems in combination with final storage in thoroughly investigated stable geologic formations. This report also briefly describes other activities carried out by the nuclear industry, namely, the construction of a central storage facility for spent fuel elements (to be in operation by 1985), a repository for reactor waste (to be in operation by 1988), and an intermediate storage facility for vitrified high-level waste (to be in operation by 1990). The R and D activities are updated to September 1981

Conceptual designs of tokamak reactor and R D

International Nuclear Information System (INIS)

Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

1983-01-01

The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

IL23R in the Swedish, Finnish, Hungarian and Italian populations: association with IBD and psoriasis, and linkage to celiac disease

Directory of Open Access Journals (Sweden)

Vatta Serena

2009-01-01

Full Text Available Abstract Background Association of the interleukin-23 receptor (IL23R with inflammatory bowel disease (IBD has been confirmed in several populations. IL23R also associates with psoriasis, suggesting that the gene may be an important candidate for many chronic inflammatory diseases. Methods We studied association of single-nucleotide variants in IL23R with IBD in Swedish patients, in both Crohn's disease (CD and ulcerative colitis (UC subsets. The same genetic variants were also studied in Finnish patients with psoriasis or celiac disease, and in Hungarian and Italian patients with celiac disease. Results Association of IL23R with IBD was replicated in our Swedish patients, and linkage and association of the IL23R region with psoriasis was found in the Finnish population. The IL23R region was also linked to celiac disease in Finnish families, but no association of IL23R variants with celiac disease was found in the Finnish, Hungarian or Italian samples. Conclusion Our study is the first to demonstrate association of IL23R with CD and UC in Swedish patients with IBD. It is also the first study to report linkage and association of the IL23R region with psoriasis in the Finnish population. Importantly, this is the first report of linkage of the IL23R region to celiac disease, a chronic inflammatory condition in which IL23R has not been previously implicated.

SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

2009-12-03

The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs

Phase transition and conduction mechanism in Pb{sub 2}Na{sub 0.8}R{sub 0.2}Nb{sub 4.8}Fe{sub 0.2}O{sub 15} material (R=rare earth)

Energy Technology Data Exchange (ETDEWEB)

Bouziane, M. [Laboratoire de Chimie du Solide Appliquée, Faculté des Sciences, Université Mohammed V-Agdal, Avenue Ibn Batouta, BP 1014 Rabat (Morocco); Taibi, M., E-mail: taibiens@yahoo.fr [Laboratoire de Physico-Chimie des Matériaux (LAF 502), Ecole Normale Supérieure, Université Mohammed V-Agdal, BP 5118 Rabat (Morocco); Boukhari, A. [Laboratoire de Chimie du Solide Appliquée, Faculté des Sciences, Université Mohammed V-Agdal, Avenue Ibn Batouta, BP 1014 Rabat (Morocco)

2013-11-15

Electrical properties of Pb{sub 2}Na{sub 0.8}Eu{sub 0.2}Nb{sub 4.8}Fe{sub 0.2}O{sub 15} tungsten bronze compound were investigated. Ferroelectric phase transition of diffuse type is observed at 395 °C. Conductivity study as a function of temperature (RT-600 °C) and at three different frequencies (10, 100 and 1000 kHz) suggests the existence of dominant ionic conduction. The rise of ac conductivity on increasing temperature supports the NTCR (negative temperature coefficient of resistance) behaviour of the material. The activation energies have been evaluated from ac conductivity using Arrhenius equation and discussed. Different conduction mechanisms were identified. For comparison, the conducting properties of Pb{sub 2}Na{sub 0.8}R{sub 0.2}Nb{sub 4.8}Fe{sub 0.2}O{sub 15} (R=Dy, Nd, La) were also investigated. - Graphical abstract: Thermal evolution of lnσ{sub ac} of Pb{sub 2}Na{sub 0.8}Eu{sub 0.2}Nb{sub 4.8}Fe{sub 0.2}O{sub 15} at selected frequencies. Display Omitted - Highlights: • We found that TB compounds exhibit a diffuse type of first- order transition. • A negative temperature coefficient of resistance (NTCR) behaviour is observed. • Three conduction mechanisms were identified: n-and/or p-type at low temperatures. • The conduction mechanism in the studied compounds is very complex.

Assessment of the impact of the Chernobyl Reactor accident on the Biota of Swedish Streams and Lakes

Energy Technology Data Exchange (ETDEWEB)

Petersen, R C; Landner, L; Blanck, H

1986-01-01

The Chernobyl reactor accident resulted in elevated levels of radionuclides in the air space above Sweden, which were then washed into Swedish lakes and streams. Before suspended particles stripped the water column, the concentration of /sup 137/Cs in small Swedish lakes was in the order of 10-40 Bq/l. This level of radioactivity should result in a negligible increase in the external exposure rate. However, by August 1986 increased levels of radioactivity were found at all trophic levels of freshwater ecosystems from algae to top carnivore, and from the available data the levels of radioactivity are still increasing. The calculated dose rate for the aquatic biota caused by the two cesium isotopes, /sup 134/Cs and /sup 137/Cs, is about 25 times higher than natural levels. While acute effectrs of the Chernobyl fallout on freshwater biota are unlikely, the long term ecological effects bear watching.

MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

Energy Technology Data Exchange (ETDEWEB)

Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

The Swedish facility for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Skoeld, K.; Capala, J. [Studsvik Medical AB (Sweden); Kierkegaard, J.; Haakansson, R. [Studsvik Nuclear AB (Sweden); Gudowska, I. [Karolinska Institute (Sweden)

2000-10-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

The Swedish facility for boron neutron capture therapy

International Nuclear Information System (INIS)

Skoeld, K.; Capala, J.; Kierkegaard, J.; Haakansson, R.; Gudowska, I.

2000-01-01

A BNCT (Boron Neutron Capture Therapy) facility has been constructed at the R2-0 reactor at Studsvik, Sweden. R2-0 is a 1 MW, open core, pool reactor. The reactor core is suspended on a movable tower and can be positioned anywhere in the pool. The BNCT facility includes two adjacent, parallel filter/moderator configurations and the reactor core is positioned in front of any of them as appropriate. One of the resulting neutron beams has been optimized for clinical irradiations with a filter/moderator system that allows easy variation of the neutron spectrum from the thermal to the epithermal energy range and with an extended collimator for convenient patient positioning. The other beam has been designed for radiobiological research and is equipped with a heavy water moderator and a large irradiation cavity with a uniform field of thermal neutrons. (author)

Insurance cost of Swedish nuclear power plants

International Nuclear Information System (INIS)

Kaellstrand, Aasa.

1992-01-01

What happens if a reactor accident occurs? Can victims of a nuclear accident be compensated for losses? The rights of a victim of a nuclear accident to be compensated for losses are governed by international conventions. These conventions make the licensee of a nuclear plant strictly liable. However, the maximum amount of compensation is limited. In Sweden the total liability of the plant-owner is maximized to 1.2 million Swedish Crowns, that is 0.02 oere/kWh. After the accidents of Harrisburg (1979) and Chernobyl (1986), it has become clear that the amounts of the various conventions are not at all sufficient to cover the damages caused by such an accident. In spite of these facts, there are a large number of reliable sources, who think that the insurance costs are negligible in the cost of production. A cost-benefit analysis based on a study performed by Ottinger et al. in 'Environmental costs of electricity' is therefore adopted to derive the costs of the external effects of nuclear plant operation and from releases to the environment during operation. The environmental externality costs of Swedish nuclear power plant operations are in this report estimated to 18.3 oere/kWh. This figure can be compared to the insurance cost, which for the present is 0.02 oere/kWh. The 'real' insurance cost including the external effects is calculated to approximately 1.12 billion Swedish Crowns] That is 900 times larger than the insurance premium, which the licensee of a nuclear plant faces] (au)

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-UO2 configuration.

Energy Technology Data Exchange (ETDEWEB)

Klann, R. T.; Perret, G.; Nuclear Engineering Division

2007-10-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R1-UO2 core configuration were completed. The reactor model was generated using the REBUS code developed at Argonne National Laboratory. The calculations are based on the specifications for fabrication, so they are considered preliminary until sampling and analysis have been completed on the fabricated samples. The estimates indicate a range of reactivity effect from -22 pcm to +25 pcm compared to the natural U sample.

The Locomotor Capabilities Index; validity and reliability of the Swedish version in adults with lower limb amputation

Directory of Open Access Journals (Sweden)

Andersson Ingemar H

2009-05-01

Full Text Available Abstract Background The Locomotor Capabilities Index (LCI is a validated measure of lower-limb amputees' ability to perform activities with prosthesis. We have developed the LCI Swedish version and evaluated its validity and reliability. Methods Cross-cultural adaptation to Swedish included forward/backward translations and field testing. The Swedish LCI was then administered to 144 amputees (55 women, mean age 74 (40–93 years, attending post-rehabilitation prosthetic training. Construct validity was assessed by examining the relationship between the LCI and Timed "Up-and-Go" (TUG test and between the LCI and EQ-5D health utility index in 2 subgroups of 40 and 20 amputees, respectively. Discriminative validity was assessed by comparing scores in different age groups and in unilateral and bilateral amputees. Test-retest reliability (1–2 weeks was evaluated in 20 amputees (14 unilateral. Results The Swedish LCI showed good construct convergent validity, with high correlation with the TUG (r = -0.75 and the EQ-5D (r = 0.84, and discriminative validity, with significantly worse mean scores for older than younger and for bilateral than unilateral amputees (p Conclusion The Swedish version of the LCI demonstrated good validity and internal consistency in adult amputees. Test-retest reliability in a small subsample appears to be acceptable. The high ceiling effect of the LCI may imply that it would be most useful in assessing amputees with low to moderate functional abilities.

SAVANNAH RIVER SITE R-REACTOR DISASSEMBLY BASIN IN-SITU DECOMMISSIONING -10499

Energy Technology Data Exchange (ETDEWEB)

Langton, C.; Serrato, M.; Blankenship, J.; Griffin, W.

2010-01-04

The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the 105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate it from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,384 cubic meters or 31,894 cubic yards. Portland cement-based structural fill materials were designed and tested for the reactor ISD project, and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and material flow considerations, maximum lift heights and differential height requirements were determined. Pertinent data and information related to the SRS 105-R Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material

Irradiation of Superheater Test Fuel Elements in the Steam Loop of the R2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Ravndal, F

1967-12-15

The design, fabrication, irradiation results, and post-irradiation examination for three superheater test fuel elements are described. During the spring of 1966 these clusters, each consisting of six fuel rods, were successfully exposed in the superheater loop No. 5 in the R2 reactor for a maximum of 24 days at a maximum outer cladding surface temperature of {approx} 650 deg C. During irradiation the linear heat rating of the rods was in the range 400-535 W/cm. The diameter of the UO{sub 2} pellets was 11.5 and 13.0 mm; the wall thickness of the 20/25 Nb and 20/35 cladding was in every case 0.4 mm. The diametrical gap between fuel and cladding was one of the main parameters and was chosen to be 0.05, 0.07 and 0.10 mm. These experiments, to be followed by one high cladding temperature irradiation ({approx} 750 deg C) and one long time irradiation ({approx} 6000 MWd/tU), were carried out to demonstrate the operational capability of short superheater test fuel rods at steady and transient operational environments for the Marviken superheater fuel elements and also to provide confirmation of design criteria for the same fuel elements.

Assessment of the Swedish EQ-5D experience-based value sets in a total hip replacement population.

Science.gov (United States)

Nemes, Szilárd; Burström, Kristina; Zethraeus, Niklas; Eneqvist, Ted; Garellick, Göran; Rolfson, Ola

2015-12-01

All patients undergoing elective total hip replacement (THR) in Sweden are asked to complete a survey, including the EQ-5D. Thus far, EQ-5D values have been presented using the UK TTO value set based on hypothetical values. Shift to the use of the recently introduced Swedish experience-based value set, derived from a representative Swedish population, is an appealing alternative. To investigate how accurate the Swedish experience-based VAS value set predicts observed EQ VAS values and to compare correlations between Swedish and UK value sets including two provisional value sets derived from the THR population. Pre- and one-year postoperative data from 56,062 THR patients from the Swedish Hip Arthroplasty Register were used. Agreement between the observed and the predicted EQ VAS values was assessed with correlation. Based on pre- and postoperative data, we constructed two provisional VAS value sets. Correlations between observed and calculated values using the Swedish VAS value set were moderate (r = 0.46) in preoperative data and high (r = 0.72) in postoperative data. Correlations between UK and register-based value sets were constantly lower compared to Swedish value sets. Register-based values and Swedish values were highly correlated. The Swedish value sets are more accurate in terms of representation of the Swedish THR patients than the currently used UK TTO value set. We find it feasible to use the experience-based Swedish value sets for further presentation of EQ-5D values in the Swedish THR population.

Imperfection detection probability at ultrasonic testing of reactor vessels

International Nuclear Information System (INIS)

Kazinczy, F. de; Koernvik, L.Aa.

1980-02-01

The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

Operating experience from Swedish nuclear power plants

International Nuclear Information System (INIS)

1999-01-01

The total production of electricity from Swedish nuclear power plants was 70.5 TWh during 1998, which is the second highest yearly production ever. Production losses due to low demand totaled 5.1 TWh combined for all twelve units and production losses due to coastdown operation totaled an additional 0.5 TWh. The reason for this low power demand was a very good supply of water to the hydropower system. Hydroelectric power production was 73.6 TWh, an increase by roughly 5 TWh since 1997. Hence, the hydroelectric power production substantially exceeded the 64 TWh expected during a normal year, i.e. a year with average rainfall. Remaining production sources, mainly fossil fuel electricity production combined with district heating, contributed with 10 TWh. The total electricity production was 154.2 TWh, the highest yearly production ever. The total electricity consumption including transmission losses was 143.5 TWh. This is also the highest consumption ever and an increase by one percent compared to 1997. The preliminary net result of the electric power trade shows a net export by 10.7 TWh. The figures above are calculated from the preliminary production results. A comprehensive report on electric power supply and consumption in Sweden is given in the 1998 Annual Report from the Swedish Power Association. Besides Oskarshamn 1, all plants have periodically been operated in load-following mode, mostly because of the abundant supply of hydropower. The energy availability for the three boiling water reactors at Forsmark averaged 93.3 % and for the three pressure water reactors at Ringhals 91.0 %, both figures are the highest ever noted. In the section 'Special Reports' three events of importance to safety that occurred during 1998 are reported. The events were all rated as level 1 according to the International Nuclear Event Scale (INES)

Characterization of 14C in Swedish light water reactors.

Science.gov (United States)

Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

2008-08-01

This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

The undersea location of the Swedish Final Repository for reactor waste, SFR - human intrusion aspects

International Nuclear Information System (INIS)

Eng, T.

1989-01-01

The Swedish Final Repository for reactor waste, SFR, is built under the Baltic sea close to the Forsmark nuclear power plant. Sixty metres of rock cover the repository caverns under the seabed. The depth of the Baltic sea is about 5-6 m at this location. A human intrusion scenario that in normal inland locations has shown to be of great importance, is a well that is drilled through or in the close vicinity of the repository. Since the land uplift in the SFR area is about 6 mm/year the undersea location of SFR ensures that no well will be drilled at this location for a considerable time while the area is covered by the Baltic sea

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

International Nuclear Information System (INIS)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

Energy Technology Data Exchange (ETDEWEB)

Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

Changes in control room at Swedish nuclear power plants; Kontrollrumsfoeraendringar vid svenska kaernkraftverk

Energy Technology Data Exchange (ETDEWEB)

Kecklund, Lena [MTO Psykologi, Huddinge (Sweden)

2005-09-15

The Swedish nuclear power plants were commissioned during a period between 1972 and 1985 and the instrumentation and control equipment are basically from that period. For several years there have been plans made for changes in all the nuclear power plants and to a certain extent the changes in control equipment and monitoring rooms have also been implemented. The object of this project was to make a comprehensive review of the changes in control room design implemented in the Swedish nuclear power plants and to describe how the MTO- (Man-Technology-Organisation) and (Man-Machine-Interface) -issues have been integrated in the process. The survey is intended to give an overall picture of the changes in control room design and man-machine-interface made in the Swedish control rooms, in order to get a deeper knowledge of the change management process and its results as well as of the management of MTO-issues in these projects. The units included in this survey are: Oskarhamn reactor 2 and 3; Ringhals reactor 2, 3 and 4; Forsmark reactor 1, 2 and 3. The Oskarshamn 1 unit has not been included in this report as it has recently undergone an extensive modernisation program as well as a detailed inspection by the SKI (Swedish Nuclear Power Inspectorate). At Ringhals 2 the modernisation work is carried out at present and the unit is also subjected to extensive inspection activities carried out by SKI and is therefore not part of this survey. This report also includes a short description of relevant standards and requirements. Then follows a presentation of the results of the plant survey, presented as case studies for three companies OKG, Ringhals and FKA. Control room changes are summarized as well as the results on specific MTO issues which has been surveyed. In all the power companies there is a joint way of working with projects concerning plant modifications. This process is described for each company separately. In the concluding of the report the strengths and

Absorption and Flux Density Measurements in an Iron Plug in R1

Energy Technology Data Exchange (ETDEWEB)

Nilsson, Ragnar; Braun, Josef

1958-11-15

Thermal, epithermal and fast neutron fluxes have been measured in a 60 cm long, 'sliced' iron plug, which has been placed in the lower iron lid of the Swedish reactor R1. Au foils, Cu foils, Mn foils, P packets, Cu wires and small Fe cylinders have been used. The gamma flux has been determined with film dosimeters. The measurements have shown that only in the first centimeters of the iron is the activation determined by the thermal flux, which decreases with a relaxation length {lambda}= (1.51 {+-} 0.02) cm. The epithermal flux is entirely predominant already after 10 cm ( {lambda} = 16 cm). The epithermal neutron flux decreases even more slowly than the fast flux ({lambda} = 6.2 cm)

Absorption and Flux Density Measurements in an Iron Plug in R1

International Nuclear Information System (INIS)

Nilsson, Ragnar; Braun, Josef

1958-11-01

Thermal, epithermal and fast neutron fluxes have been measured in a 60 cm long, 'sliced' iron plug, which has been placed in the lower iron lid of the Swedish reactor R1. Au foils, Cu foils, Mn foils, P packets, Cu wires and small Fe cylinders have been used. The gamma flux has been determined with film dosimeters. The measurements have shown that only in the first centimeters of the iron is the activation determined by the thermal flux, which decreases with a relaxation length λ= (1.51 ± 0.02) cm. The epithermal flux is entirely predominant already after 10 cm ( λ = 16 cm). The epithermal neutron flux decreases even more slowly than the fast flux (λ = 6.2 cm)

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Methyl mercury exposure in Swedish women with high fish consumption

Energy Technology Data Exchange (ETDEWEB)

Bjoernberg, Karolin Ask [Division of Metals and Health, Institute of Environmental Medicine, Karolinska Institutet, Box 210, SE-171 77, Stockholm (Sweden); Vahter, Marie [Division of Metals and Health, Institute of Environmental Medicine, Karolinska Institutet, Box 210, SE-171 77, Stockholm (Sweden); Grawe, Kierstin Petersson [Toxicology Division, National Food Administration, Box 622, SE-751 26 Uppsala (Sweden); Berglund, Marika [Division of Metals and Health, Institute of Environmental Medicine, Karolinska Institutet, Box 210, SE-171 77, Stockholm (Sweden)]. E-mail: Marika.Berglund@imm.ki.se

2005-04-01

We studied the exposure to methyl mercury (MeHg) in 127 Swedish women of childbearing age with high consumption of various types of fish, using total mercury (T-Hg) in hair and MeHg in blood as biomarkers. Fish consumption was assessed using a food frequency questionnaire (FFQ), including detailed information about consumption of different fish species, reflecting average intake during 1 year. We also determined inorganic mercury (I-Hg) in blood, and selenium (Se) in serum. The average total fish consumption, as reported in the food frequency questionnaire, was approximately 4 times/week (range 1.6-19 times/week). Fish species potentially high in MeHg, included in the Swedish dietary advisories, was consumed by 79% of the women. About 10% consumed such species more than once a week, i.e., more than what is recommended. Other fish species potentially high in MeHg, not included in the Swedish dietary advisories, was consumed by 54% of the women. Eleven percent never consumed fish species potentially high in MeHg. T-Hg in hair (median 0.70 mg/kg; range 0.08-6.6 mg/kg) was associated with MeHg in blood (median 1.7 {mu}g/L; range 0.30-14 {mu}g/L; r {sub s}=0.78; p<0.001). Hair T-Hg, blood MeHg and serum Se (median 70 {mu}g/L; range 46-154 {mu}g/L) increased with increasing total fish consumption (r {sub s}=0.32; p<0.001, r {sub s}=0.37; p<0.001 and r {sub s}=0.35; p=0.002, respectively). I-Hg in blood (median 0.24 {mu}g/L; range 0.01-1.6 {mu}g/L) increased with increasing number of dental amalgam fillings. We found no statistical significant associations between the various mercury species measured and the Se concentration in serum. Hair mercury levels exceeded the levels corresponding to the EPA reference dose (RfD) of 0.1 {mu}g MeHg/kg b.w. per day in 20% of the women. Thus, there seems to be no margin of safety for neurodevelopmental effects in fetus, for women with high fish consumption unless they decrease their intake of certain fish species.

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.

Safety and Radiation Protection at Swedish Nuclear Power Plants 2005

International Nuclear Information System (INIS)

2006-05-01

In 2005, no severe events occurred which challenged the safety at the Swedish nuclear power plants. However, some events have been given a special focus. The 'Gudrun' storm, which occurred in January 2005, affected the operation of the reactors at Ringhals and Barsebaeck 2. At Ringhals, the switchyards were affected by salt deposits and, at Barsebaeck, the 400kV grid was subjected to interruptions. The long-term trend is that the total number of fuel defects in Swedish reactors is decreasing. The damage that occurs nowadays has mainly been caused by small objects entering the fuel via the coolant and fretting holes in the cladding. To reduce the number of defects of this type, fuel with filters is successively being introduced to prevent debris from entering the fuel assemblies and cyclone filters in the facility which cleans the coolant. Since the mid-nineties, the pressurised water reactors, Ringhals 2, 3 and 4, have had problems with fuel rod bowing in excess of the safety analysis calculations. Ringhals AB (RAB) has adopted measures to rectify the bowing. Follow-up work shows that the fuel rod bowing is decreasing. The followup in 2005 of damaged tubes in the Ringhals 4 steam generators indicates a continued slow damage propagation. Tubes with defects of such a limited extent that there are adequate margins to rupture and loosening have been kept in operation. Damaged tubes with insufficient margins have plugged. During the year, previously observed minor leakage from the reactor containment in Ringhals 2 was investigated in greater detail and repaired. The investigations showed extensive corrosion attack caused by deficiencies in connection with containment construction. The ageing of electrical cables and other equipment in the I-C systems has been examined by SKI. Regulatory supervision has so far shown that these issues are largely handled in a satisfactory manner by the licensees but that certain supplementary investigations and other measures need to be

Safety and Radiation Protection at Swedish Nuclear Power Plants 2005

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-05-15

In 2005, no severe events occurred which challenged the safety at the Swedish nuclear power plants. However, some events have been given a special focus. The 'Gudrun' storm, which occurred in January 2005, affected the operation of the reactors at Ringhals and Barsebaeck 2. At Ringhals, the switchyards were affected by salt deposits and, at Barsebaeck, the 400kV grid was subjected to interruptions. The long-term trend is that the total number of fuel defects in Swedish reactors is decreasing. The damage that occurs nowadays has mainly been caused by small objects entering the fuel via the coolant and fretting holes in the cladding. To reduce the number of defects of this type, fuel with filters is successively being introduced to prevent debris from entering the fuel assemblies and cyclone filters in the facility which cleans the coolant. Since the mid-nineties, the pressurised water reactors, Ringhals 2, 3 and 4, have had problems with fuel rod bowing in excess of the safety analysis calculations. Ringhals AB (RAB) has adopted measures to rectify the bowing. Follow-up work shows that the fuel rod bowing is decreasing. The followup in 2005 of damaged tubes in the Ringhals 4 steam generators indicates a continued slow damage propagation. Tubes with defects of such a limited extent that there are adequate margins to rupture and loosening have been kept in operation. Damaged tubes with insufficient margins have plugged. During the year, previously observed minor leakage from the reactor containment in Ringhals 2 was investigated in greater detail and repaired. The investigations showed extensive corrosion attack caused by deficiencies in connection with containment construction. The ageing of electrical cables and other equipment in the I-C systems has been examined by SKI. Regulatory supervision has so far shown that these issues are largely handled in a satisfactory manner by the licensees but that certain supplementary investigations and other measures

Development Plan and R and D Status of China Lead-based Reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS [Institute of Nuclear Energy Safety Technology, Beijing (Switzerland)

2013-07-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Development Plan and R and D Status of China Lead-based Reactor

International Nuclear Information System (INIS)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS

2013-01-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

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

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)

The origin of high magnetic properties in (R,Zr)(Fe,Co)11.0–11.5Ti1.0–0.5Ny (y=1.0–1.4 for R=Nd, y=0 for R=Sm) compounds

International Nuclear Information System (INIS)

Kobayashi, K.; Suzuki, S.; Kuno, T.; Urushibata, K.; Sakuma, N.; Yano, M.; Shoji, T.; Kato, A.; Manabe, A.

2017-01-01

Ten alloys and nitrogenated compounds of (R,Zr)(Fe,Co) 11.0–11.5 Ti 1.0–0.5 N y (y=1.0–1.4 for R=Nd, y=0 for R=Sm) with a ThMn 12 -type structure were prepared. The average Fe–Fe interatomic distances, d(Fe–Fe), for Fe sites were calculated based on the reported atomic parameters. The hyperfine splittings (inner field (IF), in teslas) were measured by Mössbauer spectroscopy, and the IF increased with increasing d(Fe–Fe) for Fe sites, indicating a magneto-volume effect. The order of IF magnitude in Fe sites was Fe(8i)>Fe(8j)>Fe(8f) in all alloys. Co substitution for Fe sites, (Fe 0.75 Co 0.25 ), increased the IF by 25% for the R=Nd alloy and 15% for the R=Sm alloy. Decreasing Ti content from −Ti 1.0 to −Ti 0.5 , which increased the Fe and Co content, preserved the ThMn 12 structure with Zr substitution for R(2a) sites, and caused a slight increase in the IF of 2% for the R=Nd alloy and 7% for the R=Sm alloy. Nitrogenation, where N was introduced into the 2b sites, also increased IF in R=Nd alloys, by 23% for the Co- and Zr-free alloys, NdFe 11 Ti 1.0 N 1.5 , and by 7% for the Co-containing, (Nd 0.7 Zr 0.3 ) (Fe 0.75 Co 0.25 ) 11.5 Ti 0.5 N 1.3 alloy. The IF values of the R=Nd alloys were slightly larger than those of the R=Sm alloys. In conclusion, the magneto-volume effect was clearly observed at the Fe sites, and Co substitution into Fe sites and nitrogenation (R=Nd alloys) compensated for the increased IF. Increasing the Fe and Co fractions also increased IF slightly. - Highlights: • Average distances of Fe–Fe (d(Fe–Fe)) were calculated using lattice constants. • Hyperfine fields (IF) in Fe sites were measured using Mössbauer spectroscopy. • Relationship between d(Fe–Fe) and IF at each Fe site was obtained. • Co substitution and N introduction effects on IF was also measured. • Magneto-volume effect is main reason of IF augmentation in Fe sites.

A swedish dose passport - contractors point of view

International Nuclear Information System (INIS)

Andersson, M.; Holmqvist, A.; Moller, J.

2003-01-01

Westinghouse Atom is situated in Vasteras approximately 100 km west from Stockholm. The company is owned by BNFL. The two largest divisions are the Nuclear Fuel Operations and The Global Reactor Services division. The Nuclear fuel operations manufacture fuel for BWR and PWR reactors. The raw material used is Uranium hexafluoride, which is converted to Uranium dioxide powder through wet AUC-process. The concession is 600 tonnes of UO 2 , per year. Last year the production. was approximately 900 fuel elements. There is also a control rod production line within the fuel factory. Last year the production of control rods was approximately 160. The Global Reactor Services Division performs tests on different types of equipments used in nuclear power plants. In addition there is also a well-established service structure that provides a wide range of field services, for instance sipping of fuel elements. The total amount of people working in Vasteras is currently around 800. The majority of those, work at the fuel factory. The purpose of this paper is to describe the somewhat awkward situation for our employees when working as external personnel on German nuclear installations. Our Swedish personnel are currently using German dose passports. Since Sweden joined the European Union in 1995 this is in contradiction to the EU-directives. Hence, Westinghouse Atom has applied for a license for the use of Swedish dose passports in Germany. The amount of people performing service jobs in Germany is approximately 80 persons. (authors)

Radioactive inventory in structural materials of ET-R R-1 reactor and its implication on decommissioning.

Energy Technology Data Exchange (ETDEWEB)

Elkady, A; Amin, E [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

1995-10-01

A plan for decommissioning of ET-R R-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences decommissioning. Conservative calculations have been made to evaluate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are present in significant quantities in the reactor structural materials are aluminium, 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 Co-60 and Fe-55 which are present in aluminium as trace elements in larger quantities in other construction materials. 2 figs., 4 tabs.

RA-0 reactor. New neutronic calculations

International Nuclear Information System (INIS)

Rumis, D.; Leszczynski, F.

1990-01-01

An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author) [es

Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

Energy Technology Data Exchange (ETDEWEB)

Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

2009-06-15

ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

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)

Lead-based Fast Reactor Development Plan and R&D Status in China

International Nuclear Information System (INIS)

Wu Yican

2013-01-01

• Lead-based fast reactors have good potential for waste transmutation, fuel breeding and energy production, which has been selected by CAS as the advanced reactor development emphasis with the support of ADS program and MFE program. Sharing of technologies R&D is possible among GIF/ADS/Fusion. • The concepts and test strategy of series China lead-based fast reactors (CLEAR) have been developed. The preliminary engineering design and safety analysis of CLEAR-I are underway. • Technology R&D on CLEAR with series lead alloy loops and accelerator-based neutron generator have been constructed or under construction. • CLEAR series reactor design and construction have big challenges, widely international cooperation on reactor design and technology R&D is welcome

Synthesis of carbasugars from aldonolactones, part III - A study on the allylic substitution of (1R,5R,8R)- and (1R,5R,8S)-8-hydroxy-2-oxabicyclo[3.3.0]oct-6-en-3-one derivatives - Preparation of (1S,2R,3R)-9-[2-hydroxy-3-(2-hydroxyethyl)cyclopent-4-en-1-yl]-9H-adenine

DEFF Research Database (Denmark)

Johansen, Steen Karsk; Lundt, Inge

2001-01-01

The palladium-catalyzed substitution of acylated (1R,5R,8R)- and (1R,SR,8S)-8-hydroxy-2-oxabicyclo[3.3.0] ones has been studied using a number of C- and N-nucleophiles, In all cases, the exo derivatives (8R) were found to be more reactive than the corresponding endo derivatives (8S). The reaction...... with these nucleophiles. Additionally, Mitsunobu substitution of (1R,5R,8R)-8-hydroxy-2-oxabicyclo[3.3.0]oct-B-en-3-one (3) with 6-chloropurine, followed by reduction of the lactone moiety and treatment with Liquid ammonia, gave the carbocyclic nucleoside (1S,2R,3R)-9-[2-hydroxy-3-(2-hydroxyethyl)cyclopent-4-en-1-yl]-9H...

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Energy Technology Data Exchange (ETDEWEB)

Wiersma, B.

2010-05-24

operations in the R-reactor vessel is low for the Portland cement. Alternatively, if the grout fill rate is less than 0.5 inch/min and the grout is maintained at a temperature of 80 C, the risk is again low. Although these calculations are conservative, there are some measures that may be taken to further minimize the potential for hydrogen evolution. (1) Minimize the temperature of the grout as much as practical. Lower temperatures will mean lower hydrogen generation rates. For P-reactor, grout temperatures less than 100 C should provide an adequate safety margin for the pH 8 and pH 10.4 grout formulations. For R-reactor, grout temperatures less than 70 C or 80 C will provide an adequate safety margin for the Portland cement. The other grout formulations are also viable options for R-reactor. (2) Minimize the grout fill rate as much as practical. Lowering the fill rate takes advantage of passivation of the aluminum components and hence lower hydrogen generation rates. For P-reactor, fill rates that are less than 2 inches/min for the ceramicrete and the silica fume grouts will reduce the chance of significant hydrogen accumulation. For R-reactor, fill rates less than 1 inch/min will again minimize the risk of hydrogen accumulation. (3) Ventilate the building as much as practical (e.g., leave doors open) to further disperse hydrogen. The volumetric hydrogen generation rates in the P-reactor vessel, however, are low for the pH 8 and pH 10.4 grout, (i.e., less than 0.97 ft{sup 3}/min). If further walk-down inspections of the reactor vessels suggest an increase in the actual areal density of aluminum, the calculations should be re-visited.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

How to interpret Swedish energy policy - Facts and analysis

International Nuclear Information System (INIS)

Rising, Agneta; Bohl, Torsten; Wikdahl, Carl-Erik

1998-01-01

The Swedish parliament decided on June 10, 1997 that one of the two reactors at the Barsebaeck nuclear power plant shall be closed before mid 1998 and the other until three years later. Some weeks before the 1998 PIME Conference (on December 18) the same parliament is planning to accept a new act, which will make it possible for the government to close any reactor in the future without ay reference to the level of safety. Sweden is known 'internationally to have a successful nuclear power programme and to be in the front line to develop safe nuclear waste methods. The decision in the Swedish parliament therefore came as a surprise not only in Sweden but to a large part of the nuclear power industry, all over the world. Nuclear power accounts for half the power generated in Sweden. here are twelve nuclear power units with a net output of 10 000 MW and an annual energy generation capacity of more than 70 TWh. Nuclear production in Sweden has proved to be technically, economically and environmentally highly successful. ne capacity factors have normally been high, the production costs are low and so are the releases of radioactivity and doses to the personnel. All twelve nuclear units are still highly competitive generators on the deregulated Nordic electricity market and a fe time of at least 40 years is expected for a the nuclear units, as they are being modernised continuously. The estimated safety standard of all twelve units is among the highest in the world. A dynamic nuclear waste programme has been launched. Swedish waste management techniques have achieved world leadership in several important areas. The main part of the explanation can be found in the skilful political strategy of one or two political parties which have been advocating the premature phase-out of the nuclear power programme since the mid 70's. The anti- nuclear policy was introduced in the Swedish parliament already in the 1976 general election, when the Centre Party with a strong antinuclear

Accident dynamics of LR-0 reactor

International Nuclear Information System (INIS)

Vorisek, M.; Tinka, I.

1981-01-01

The results are given of calculating the accident dynamics of the LR-0 light water experimental zero power reactor. Calculations of the time dependence of power, the total released energy, the temperature of fuel and its cladding were made using program FATRAP for different values of the total inserted reactivity. Using the results, an analysis is made of hypothetic accident states of the LR-0 reactor. The results are shown graphically. (J.B.)

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)

Summary of operational experience in Swedish nuclear power plants 1995

International Nuclear Information System (INIS)

1996-01-01

A summary of two pages for each Swedish reactor is given with availability, number of scrams, collective radiation doses and events for 1995. Special reports are presented on some specific issues: Bowed fuel assemblies at Ringhals, Incorrect opening pressure of the main safety valves at Ringhals, Measures to restore and upgrade safety at Oskarshamn 1, and the Decontamination of the reactor vessel at Oskarshamn 1. Figs

The Swedish Utilities joint approach to form common basis for design requirements for the future

International Nuclear Information System (INIS)

Hansson, B.

1998-01-01

The Owners of the Swedish Nuclear Power Plants have decided to form a document that should state the design principals and requirement for cost-effective and continuous development of the reactor safety in the future. The development of this document will be a part of the modernization and development of the Swedish Nuclear Power Plants. The basis for this document is an evaluation of Swedish and International standards and regulations as IAEA/INSAG, US-regulations, EUR etc. (author)

R and D directions for the development of CANDU reactors

International Nuclear Information System (INIS)

Torgerson, D.F.

1998-01-01

Full text: AECL is carrying out a comprehensive R and D programme to advance all aspects of CANDU reactor technology. These programs are focusing on three main strategic directions: improved economics, enhanced safety, and fuel cycle flexibility. R and D areas include fuel cycle development, heavy water technology, fuel channel development, safety technology, control and instrumentation, reactor chemistry, systems and components, and health and environment. In each case, the R and D programs have short, medium, and long-term goals to achieve the overall strategic directions. Most of the programs seek to further develop and exploit some of the unique characteristics of pressurized heavy water reactors. Examples of this include high neutron economy and on-power fueling which allow several different fuel cycles, the presence of large water heat sinks for enhanced safety, and modular components that can be easily replaced for plant life extension. This presentation reviews AECL's product development directions and the R and D programs that have been begun for their development

Operating experience from Swedish nuclear power plants, 2000

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

From a safety point of view, 2000 was - as were previous years - satisfactory. Total electricity production from the Swedish nuclear power stations amounted to 54.2 TWh, which was over 20% less than the 70.2 TWh produced in 1999. The two main reasons for the reduction were the closure of Barsebaeck 1 on 1st December 1999, and the cutback in output from all reactors due to the particularly good availability of hydro power in 2000. Some reactors were even shut down completely as a result of the low power demand, which has not happened previously. The quantity of unutilised production capacity as a result of these reductions amounted to 11.6 TWh. Costdown operation prior to the annual overhaul shutdowns, which makes better use of the fuel, represented a further 2.1 TWh of unutilised capacity. The average energy availability of the three PWRs at Ringhals was 82.0%, while that of the eight BWRs was 84.2%. Forsmark 3, Ringhals 3 and Oskarshamn 3 all had average availabilities of over 90%. Of five events with safety implications that occurred in the plants during the year, three are described under Special Reporting. One of them relates to the crack indications in welds that were found in an American PWR in the autumn, and which were subsequently also found in Ringhals 4.

Operating experience from Swedish nuclear power plants, 2000

International Nuclear Information System (INIS)

2001-01-01

From a safety point of view, 2000 was - as were previous years - satisfactory. Total electricity production from the Swedish nuclear power stations amounted to 54.2 TWh, which was over 20% less than the 70.2 TWh produced in 1999. The two main reasons for the reduction were the closure of Barsebaeck 1 on 1st December 1999, and the cutback in output from all reactors due to the particularly good availability of hydro power in 2000. Some reactors were even shut down completely as a result of the low power demand, which has not happened previously. The quantity of unutilised production capacity as a result of these reductions amounted to 11.6 TWh. Costdown operation prior to the annual overhaul shutdowns, which makes better use of the fuel, represented a further 2.1 TWh of unutilised capacity. The average energy availability of the three PWRs at Ringhals was 82.0%, while that of the eight BWRs was 84.2%. Forsmark 3, Ringhals 3 and Oskarshamn 3 all had average availabilities of over 90%. Of five events with safety implications that occurred in the plants during the year, three are described under Special Reporting. One of them relates to the crack indications in welds that were found in an American PWR in the autumn, and which were subsequently also found in Ringhals 4

Improving EGSB reactor performance for simultaneous bioenergy and organic acid production from cheese whey via continuous biological H2 production.

Science.gov (United States)

Ramos, Lucas Rodrigues; Silva, Edson Luiz

2017-07-01

To evaluate the influence of hydraulic retention time (HRT) and cheese whey (CW) substrate concentration (15 and 25 g lactose l -1 ) on the performance of EGSB reactors (R15 and R25, respectively) for H 2 production. A decrease in the HRT from 8 to 4 h favored the H 2 yield and H 2 production rate (HPR) in R15, with maximum values of 0.86 ± 0.11 mmol H 2 g COD -1 and 0.23 ± 0.024 l H 2 h -1 l -1 , respectively. H 2 production in R25 was also favored at a HRT of 4 h, with maximum yield and HPR values of 0.64 ± 0.023 mmol H 2 g COD -1 and 0.31 ± 0.032 l H 2 h -1 l -1 , respectively. The main metabolites produced were butyric, acetic and lactic acids. The EGSB reactor was evaluated as a viable acidogenic step in the two-stage anaerobic treatment of CW for the increase of COD removal efficiency and biomethane production.

Shutting down two reactors

International Nuclear Information System (INIS)

1987-12-01

Nuclear power will be phased out of the swedish energy system during the first decades of the next century. Commissioned by the swedish government, the National Energy Administration reports a study on the possibilities for, and consequences of, an earlier shut down (1994-1996) of two of the twelve swedish power reactors. Some of the questions studied are: How much will the electricity price raise ?; How will the electricity consumption be affected ?; What are the alternatives to nuclear power ?; What will the cost be ? and What will the environmental effects be ?. (L.E.)

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

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)

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)

Safety considerations concerning light water reactors in Sweden

International Nuclear Information System (INIS)

Nilsson, T.

1977-01-01

In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

Report on the status of instrumentation and control in Swedish nuclear power plants

International Nuclear Information System (INIS)

Stroebeck, E.

1992-01-01

Nuclear power plants accounted for 46% of the total electric power production in Sweden in 1990. The availability of the Swedish reactors remains at a very high level. The oldest Swedish nuclear power plant has been in operation for nearly 20 years, and in the next 5 to 10 years a large portion of the NPP electrical equipment has to be replaced. The paper presents an overview of activities on control and instrumentation in the following: Future developments; implementation of computer-based systems; training simulators; nuclear safety research. The operating experience in Swedish nuclear power plants in 1991 is also presented. (author)

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.)

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)

Reactor safety research in Sweden

International Nuclear Information System (INIS)

Pershagen, B.

1980-02-01

Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

Thermal expansion of phosphates with the NaZr2(PO4)3 structure containing lanthanides and zirconium: R0.33Zr2(PO4)3 (R = Nd, Eu, Er) and Er0.33(1–x) Zr0.25xZr2(PO4)3

International Nuclear Information System (INIS)

Volgutov, V. Yu.; Orlova, A. I.

2015-01-01

Phosphates R 0.33 Zr 2 (PO 4 ) 3 (R = Nd, Eu, or Er) and Er 0.33(1–x) Zr 0.25 Zr 2 (PO 4 ) 3 (x = 0, 0.25, 0.5, 0.75, 1.0) of the NaZr 2 (PO 4 ) 3 family have been synthesized and investigated by high-temperature X-ray diffraction. The crystallochemical approach is used to obtain compounds with expected small and controllable thermal-expansion parameters. Phosphates with close-to-zero thermal-expansion parameters, including those with low thermal-expansion anisotropy, have been obtained: Nd 0.33 Zr 2 (PO 4 ) 3 with α a =–2.21 × 10 −6 °C −1 , α c = 0.81 × 10 −6 °C −1 , and Δα = 3.02 × 10 −6 °C –1 and Er 0.08 Zr 0.19 Zr 2 (PO 4 ) 3 with α a =–1.86 × 10 −6 °C −1 , α c = 1.73 × 10 −6 °C −1 , and Δα = 3.58 × 10 −6 °C −1

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)

Swedish spent fuel management systems, facilities and operating experiences

International Nuclear Information System (INIS)

Vogt, J.

1998-01-01

About 50% of the electricity in Sweden is generated by means of nuclear power from 12 LWR reactors located at four sites and with a total capacity of 10,000 MW. The four utilities have jointly created SKB, the Swedish Nuclear Fuel and Waste Management Company, which has been given the mandate to manage the spent fuel and radioactive waste from its origin at the reactors to the final disposal. SKB has developed a system for the safe handling of all kinds of radioactive waste from the Swedish nuclear power plants. The keystones now in operation of this system are a transport system, a central interim storage facility for spent nuclear fuel (CLAB), a final repository for short-lived, low and intermediate level waste (SFR). The remaining, system components being planned are an encapsulation plant for spent nuclear fuel and a deep repository for encapsulated spent fuel and other long-lived radioactive wastes. (author)

The molten salt reactor: R and D status and perspectives in Europe

International Nuclear Information System (INIS)

Renault, Claude; Delpech, Sylvie; Merle-Lucotte, Elsa; Konings, Rudy; Hron, Miloslav; Ignatiev, Victor

2010-01-01

The paper concentrates on molten salt fast reactor (MSFR) concepts which are receiving most attention in the EU context. It shows the main R and D achievements and some remaining issues to be addressed in such essential areas as (a) reactor conceptual design, (b) molten salt properties, (c) fuel salt clean-up scheme and (d) high temperature materials. The status and perspectives of molten salt reactor R and D efforts in Europe are then discussed

Neutron flux of 100kW in the irradiation terminals of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante Marco

2009-01-01

In this work, it was carried out a study of the neutron flux in the IPR-R1 TRIGA reactor irradiation facilities: rotary specimen rack (RSR), pneumatic transfer tube two (PTT2) and the central thimble (CT). The objective was to obtain the neutron flux profile on the RSR, which has forty irradiation positions, and also values for the thermal and epithermal neutron fluxes of some RSR positions and also of the PTT2 and of the CT facility. It was applied the neutron activation analysis of a reference material, Al-Au (0.1%) alloy. Irradiations were performed on 16 different dates. It was concluded that for the RSR, the average value of thermal and epithermal neutron fluxes depends on the vertical position of the reactor control rods. Neutron flux variations along the RSR form a characteristic profile, whose values depend on the location of the irradiation position in the reactor core and on the control rods vertical position. In the RSR, the obtained values of thermal and epithermal neutron flux were (8.1 +- 0.3) x 10 11 n.cm -2 .s -1 , and (3.4 +- 0.2)x10 10 n.cm -2 .s -1 , respectively. For the PTT2 and the CT, the values for the epithermal neutron flux were respectively (3.3 +- 0.2) x 10 9 n.cm -2 .s -1 and (2.6 +- 0.1) x 10 11 n.cm -2 .s -1 . For these facilities, the thermal neutron flux was estimated, and the obtained values were (2.4 +- 0.2) x 10 11 n.cm -2 .s -1 and (2.8 +- 0.1)x10 12 n.cm -2 .s -1 for the PTT2 and the CT, respectively. (author)

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

Quarterly report of the Swedish Nuclear Power Inspectorate

International Nuclear Information System (INIS)

1984-01-01

The inspectorate is reporting on the departures of the nuclear power plants from normal operations. No safety incidents of importance occurred during the 4th quarter 1983. There have been 12 reactor trips for the 10 power units, 9 of those occurred on Dec. 27th, when the Southern Swedish power grid tripped.(P.Aa.)

Characteristics of surface acoustic waves in (11\\bar 2 0)ZnO film/ R-sapphire substrate structures

Science.gov (United States)

Wang, Yan; Zhang, ShuYi; Xu, Jing; Xie, YingCai; Lan, XiaoDong

2018-02-01

(11\\bar 2 0)ZnO film/ R-sapphire substrate structure is promising for high frequency acoustic wave devices. The propagation characteristics of SAWs, including the Rayleigh waves along [0001] direction and Love waves along [1ī00] direction, are investigated by using 3 dimensional finite element method (3D-FEM). The phase velocity ( v p), electromechanical coupling coefficient ( k 2), temperature coefficient of frequency ( TCF) and reflection coefficient ( r) of Rayleigh wave and Love wave devices are theoretically analyzed. Furthermore, the influences of ZnO films with different crystal orientation on SAW properties are also investigated. The results show that the 1st Rayleigh wave has an exceedingly large k 2 of 4.95% in (90°, 90°, 0°) (11\\bar 2 0)ZnO film/ R-sapphire substrate associated with a phase velocity of 5300 m/s; and the 0th Love wave in (0°, 90°, 0°) (11\\bar 2 0)ZnO film/ R-sapphire substrate has a maximum k 2 of 3.86% associated with a phase velocity of 3400 m/s. And (11\\bar 2 0)ZnO film/ R-sapphire substrate structures can be used to design temperature-compensated and wide-band SAW devices. All of the results indicate that the performances of SAW devices can be optimized by suitably selecting ZnO films with different thickness and crystal orientations deposited on R-sapphire substrates.

The origin of high magnetic properties in (R,Zr)(Fe,Co){sub 11.0–11.5}Ti{sub 1.0–0.5}N{sub y} (y=1.0–1.4 for R=Nd, y=0 for R=Sm) compounds

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, K., E-mail: koba@ms.sist.ac.jp [Shizuoka Institute of Science and Technology, 2200-2 Toyosawa, Fukuroi, Shizuoka 437-8555 (Japan); Suzuki, S.; Kuno, T.; Urushibata, K. [Shizuoka Institute of Science and Technology, 2200-2 Toyosawa, Fukuroi, Shizuoka 437-8555 (Japan); Sakuma, N.; Yano, M.; Shoji, T.; Kato, A. [Toyota Motor Corporation, 1200 Mishuku, Susono, Shizuoka 410-1193 (Japan); Technology Reserch Association of Magnetic Materials fot High-efficiency Motors (Mag-HEM) Higashifuji-Branch, 1200 Mishuku, Susono, Shizuoka 410-1193, Japan. (Japan); Manabe, A. [Technology Reserch Association of Magnetic Materials fot High-efficiency Motors (Mag-HEM) Higashifuji-Branch, 1200 Mishuku, Susono, Shizuoka 410-1193, Japan. (Japan)

2017-03-15

Ten alloys and nitrogenated compounds of (R,Zr)(Fe,Co){sub 11.0–11.5}Ti{sub 1.0–0.5}N{sub y} (y=1.0–1.4 for R=Nd, y=0 for R=Sm) with a ThMn{sub 12}-type structure were prepared. The average Fe–Fe interatomic distances, d(Fe–Fe), for Fe sites were calculated based on the reported atomic parameters. The hyperfine splittings (inner field (IF), in teslas) were measured by Mössbauer spectroscopy, and the IF increased with increasing d(Fe–Fe) for Fe sites, indicating a magneto-volume effect. The order of IF magnitude in Fe sites was Fe(8i)>Fe(8j)>Fe(8f) in all alloys. Co substitution for Fe sites, (Fe{sub 0.75}Co{sub 0.25}), increased the IF by 25% for the R=Nd alloy and 15% for the R=Sm alloy. Decreasing Ti content from −Ti{sub 1.0} to −Ti{sub 0.5}, which increased the Fe and Co content, preserved the ThMn{sub 12} structure with Zr substitution for R(2a) sites, and caused a slight increase in the IF of 2% for the R=Nd alloy and 7% for the R=Sm alloy. Nitrogenation, where N was introduced into the 2b sites, also increased IF in R=Nd alloys, by 23% for the Co- and Zr-free alloys, NdFe{sub 11}Ti{sub 1.0}N{sub 1.5}, and by 7% for the Co-containing, (Nd{sub 0.7}Zr{sub 0.3}) (Fe{sub 0.75}Co{sub 0.25}) {sub 11.5}Ti{sub 0.5}N{sub 1.3} alloy. The IF values of the R=Nd alloys were slightly larger than those of the R=Sm alloys. In conclusion, the magneto-volume effect was clearly observed at the Fe sites, and Co substitution into Fe sites and nitrogenation (R=Nd alloys) compensated for the increased IF. Increasing the Fe and Co fractions also increased IF slightly. - Highlights: • Average distances of Fe–Fe (d(Fe–Fe)) were calculated using lattice constants. • Hyperfine fields (IF) in Fe sites were measured using Mössbauer spectroscopy. • Relationship between d(Fe–Fe) and IF at each Fe site was obtained. • Co substitution and N introduction effects on IF was also measured. • Magneto-volume effect is main reason of IF augmentation in Fe sites.

Operating experience from Swedish nuclear power plants 2004

International Nuclear Information System (INIS)

2005-01-01

2004 was somewhat of a record year for the Swedish nuclear power stations. No serious faults occurred, and production exceeded previous record outputs. Total output from the eleven nuclear power units during the year amounted to 75 TWh, which is the largest amount of power ever produced by nuclear power in Sweden. Corresponding figures for earlier years are 59 TWh (2003), 65 TWh (2002) and 69 TWh (2001). An important reason for this excellent result was the very high energy availability. Forsmark 1, for example, exceeded 97 % availability, while Forsmark 2 just reached 97 %. For all the Swedish nuclear power stations as a whole, availability in 2004 amounted to 91 %. In addition to the connection between production and energy availability, there is also a connection with safety. During the year, safety in the Swedish power stations has been high, not only in absolute terms but also in an international perspective. One measure of safety is to be found in the number of accidents, incidents, anomalies or deviations reported to the IAEA on a scale known as the International Nuclear Event Scale (INES). Sweden has undertaken to report all events in accordance with this international system. Three reports were submitted by the Swedish Nuclear Power Inspectorate, which is responsible for national reporting, during the year. None of them had any significance for reactor safety: all were categorised as incidents or minor deviations from the regulations. Summarising, 2004 has been an excellent year for nuclear power safety, which is also reflected by the record electricity production during the year

Operating experience from Swedish nuclear power plants 2004

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

2004 was somewhat of a record year for the Swedish nuclear power stations. No serious faults occurred, and production exceeded previous record outputs. Total output from the eleven nuclear power units during the year amounted to 75 TWh, which is the largest amount of power ever produced by nuclear power in Sweden. Corresponding figures for earlier years are 59 TWh (2003), 65 TWh (2002) and 69 TWh (2001). An important reason for this excellent result was the very high energy availability. Forsmark 1, for example, exceeded 97 % availability, while Forsmark 2 just reached 97 %. For all the Swedish nuclear power stations as a whole, availability in 2004 amounted to 91 %. In addition to the connection between production and energy availability, there is also a connection with safety. During the year, safety in the Swedish power stations has been high, not only in absolute terms but also in an international perspective. One measure of safety is to be found in the number of accidents, incidents, anomalies or deviations reported to the IAEA on a scale known as the International Nuclear Event Scale (INES). Sweden has undertaken to report all events in accordance with this international system. Three reports were submitted by the Swedish Nuclear Power Inspectorate, which is responsible for national reporting, during the year. None of them had any significance for reactor safety: all were categorised as incidents or minor deviations from the regulations. Summarising, 2004 has been an excellent year for nuclear power safety, which is also reflected by the record electricity production during the year.

Application of the k{sub 0}-INAA method for analysis of biological samples at the pneumatic station of the IEA-R1 nuclear research reactor

Energy Technology Data Exchange (ETDEWEB)

Puerta, Daniel C.; Figueiredo, Ana Maria G.; Semmler, Renato, E-mail: dcpuerta@hotmail.com, E-mail: anamaria@ipen.br, E-mail: rsemmler@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Jacimovic, Radojko, E-mail: radojko.jacimovic@ijs.si [Jozef Stefan Institute (JSI), Ljubljana, LJU (Slovenia). Department of Environmental Sciences

2013-07-01

As part of the process of implementation of the k{sub 0}-INAA standardization method at the Neutron Activation Laboratory (LAN-IPEN), Sao Paulo, Brazil, this study presents the results obtained for the analysis of short and medium-lived nuclides in biological samples by k{sub 0}-INAA using the program k{sub 0}-IAEA, provided by the International Atomic Energy Agency (IAEA). The elements Al, Ba, Br, Na, K, Mn, Mg, Sr and V were determined with respect to gold ({sup 197}Au) using the pneumatic station facility of the IEA-R1 4.5 MW swimming pool nuclear research reactor, Sao Paulo. Characterization of the pneumatic station was carried out by using the 'bare triple-monitor' method with {sup 197}Au-{sup 96}Zr-{sup 94}Zr. The Certified Reference Material IRMM-530R Al-0.1%Au alloy and high purity zirconium comparators were used. The efficiency curves of the gamma-ray spectrometer used were determined by measuring calibrated radioactive sources at the usually utilized counting geometries. The method was validated by analyzing the reference materials NIST SRM 1547 Peach Leaves, INCT-MPH-2 Mixed Polish Herbs and NIST SRM 1573a Tomato Leaves. The concentration results obtained agreed with certified, reference and recommended values, showing relative errors (bias, %) less than 30% for most elements. The Coefficients of Variation were below 20%, showing a good reproducibility of the results. The E{sub n}-number showed that all results, except Na in NIST SRM 1547 and NIST SRM 1573a and Al in INCT-MPH-2, were within 95% confidence interval. (author)

Formation of the high-spin Hf-179m2 isomer in reactor irradiations

Czech Academy of Sciences Publication Activity Database

Karamian, S. A.; Carroll, J. J.; Adam, Jindřich; Kulagin, EN.; Shabalin, EP.

2004-01-01

Roč. 14, č. 4 (2004), s. 438-441 ISSN 1054-660X R&D Projects: GA MŠk(CZ) ME 134 Keywords : reactor irradiation * high-spin Hf-179m2 Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.836, year: 2004

Operating experience 1993 in Swedish nuclear power plants

International Nuclear Information System (INIS)

1995-01-01

For many years, the Swedish nuclear power plants had a very good track record, compared with the international average. This trend was broken in 1993. During the year, six power plants were shut down for extended periods of time, for different safety-related reasons. During the autumn, a reactor containment leak was detected during scheduled containment leak rate testing at Barsebaeck 2. The unit was shut down for extensive investigation and corrective action for the rest of the year. Ringhals 2 was shut down last six months of the year as crack indications were found in a weld next to a control rod penetration in the reactor vessel head. Extensive tests and analyses revealed that the crack originated from the manufacturing of the vessel head and was of minor importance to safety. Oskarshamn 1 was shut down the whole year. Cracks in cold bent pipes in the residual heat removal system and cracks in the feedwater riser pipes lead to extensive replacement of piping, including pipes inside the reactor vessel. Decontamination of the reactor vessel was successful and attracted world wide interest. A programme for plant status verification was started in order to establish long-term operating conditions. Replacement of the pipe insulation and the inlet strainers in the core and containment spray systems solved the problems with clogging at certain failures in Barsebaeck, Ringhals 1 and Oskarshamn 1 and 2. Six of the reactors had an extremely high availability, of about 90 per cent and more. By year end, eleven of the twelve reactors were in full power operation

Genetic anticipation in Swedish Lynch syndrome families.

Science.gov (United States)

von Salomé, Jenny; Boonstra, Philip S; Karimi, Masoud; Silander, Gustav; Stenmark-Askmalm, Marie; Gebre-Medhin, Samuel; Aravidis, Christos; Nilbert, Mef; Lindblom, Annika; Lagerstedt-Robinson, Kristina

2017-10-01

Among hereditary colorectal cancer predisposing syndromes, Lynch syndrome (LS) caused by mutations in DNA mismatch repair genes MLH1, MSH2, MSH6 or PMS2 is the most common. Patients with LS have an increased risk of early onset colon and endometrial cancer, but also other tumors that generally have an earlier onset compared to the general population. However, age at first primary cancer varies within families and genetic anticipation, i.e. decreasing age at onset in successive generations, has been suggested in LS. Anticipation is a well-known phenomenon in e.g neurodegenerative diseases and several reports have studied anticipation in heritable cancer. The purpose of this study is to determine whether anticipation can be shown in a nationwide cohort of Swedish LS families referred to the regional departments of clinical genetics in Lund, Stockholm, Linköping, Uppsala and Umeå between the years 1990-2013. We analyzed a homogenous group of mutation carriers, utilizing information from both affected and non-affected family members. In total, 239 families with a mismatch repair gene mutation (96 MLH1 families, 90 MSH2 families including one family with an EPCAM-MSH2 deletion, 39 MSH6 families, 12 PMS2 families, and 2 MLH1+PMS2 families) comprising 1028 at-risk carriers were identified among the Swedish LS families, of which 1003 mutation carriers had available follow-up information and could be included in the study. Using a normal random effects model (NREM) we estimate a 2.1 year decrease in age of diagnosis per generation. An alternative analysis using a mixed-effects Cox proportional hazards model (COX-R) estimates a hazard ratio of exp(0.171), or about 1.19, for age of diagnosis between consecutive generations. LS-associated gene-specific anticipation effects are evident for MSH2 (2.6 years/generation for NREM and hazard ratio of 1.33 for COX-R) and PMS2 (7.3 years/generation and hazard ratio of 1.86). The estimated anticipation effects for MLH1 and MSH6 are

Design of first reactor protection system prototype for C A R E M reactor

International Nuclear Information System (INIS)

Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

2006-01-01

In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

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)

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

AFCI-2.0 Neutron Cross Section Covariance Library

Energy Technology Data Exchange (ETDEWEB)

Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-03-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

AFCI-2.0 Neutron Cross Section Covariance Library

International Nuclear Information System (INIS)

Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-01-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

MinION Analysis and Reference Consortium: Phase 2 data release and analysis of R9.0 chemistry.

Science.gov (United States)

Jain, Miten; Tyson, John R; Loose, Matthew; Ip, Camilla L C; Eccles, David A; O'Grady, Justin; Malla, Sunir; Leggett, Richard M; Wallerman, Ola; Jansen, Hans J; Zalunin, Vadim; Birney, Ewan; Brown, Bonnie L; Snutch, Terrance P; Olsen, Hugh E

2017-01-01

Long-read sequencing is rapidly evolving and reshaping the suite of opportunities for genomic analysis. For the MinION in particular, as both the platform and chemistry develop, the user community requires reference data to set performance expectations and maximally exploit third-generation sequencing. We performed an analysis of MinION data derived from whole genome sequencing of Escherichia coli K-12 using the R9.0 chemistry, comparing the results with the older R7.3 chemistry. We computed the error-rate estimates for insertions, deletions, and mismatches in MinION reads. Run-time characteristics of the flow cell and run scripts for R9.0 were similar to those observed for R7.3 chemistry, but with an 8-fold increase in bases per second (from 30 bps in R7.3 and SQK-MAP005 library preparation, to 250 bps in R9.0) processed by individual nanopores, and less drop-off in yield over time. The 2-dimensional ("2D") N50 read length was unchanged from the prior chemistry. Using the proportion of alignable reads as a measure of base-call accuracy, 99.9% of "pass" template reads from 1-dimensional ("1D")  experiments were mappable and ~97% from 2D experiments. The median identity of reads was ~89% for 1D and ~94% for 2D experiments. The total error rate (miscall + insertion + deletion ) decreased for 2D "pass" reads from 9.1% in R7.3 to 7.5% in R9.0 and for template "pass" reads from 26.7% in R7.3 to 14.5% in R9.0. These Phase 2 MinION experiments serve as a baseline by providing estimates for read quality, throughput, and mappability. The datasets further enable the development of bioinformatic tools tailored to the new R9.0 chemistry and the design of novel biological applications for this technology. K: thousand, Kb: kilobase (one thousand base pairs), M: million, Mb: megabase (one million base pairs), Gb: gigabase (one billion base pairs).

Water scrubbers as new mitigating devices in Swedish reactors

International Nuclear Information System (INIS)

Espefaelt, R.

1988-01-01

Controlling the containment pressure is an important part of the Swedish severe accident mitigation strategy. As a final measure, venting of the containment atmosphere to the environment is feasible via a filtered venting system using a water scrubber as the filtering device. The comprehensive theoretical and experimental verification of the Multi Venturi Scrubber System has resulted in the following predicted scrubber performance: Both during the scrubber heat-up phase and in long periods of operation, where the water of the scrubber is heated to saturation, a decontamination factor of the order of several thousand is predicted. During no conditions foreseen in the safety analysis are decontamination factors below DF = 500 in the BWR scrubber and DF = 1500 in the PWR scrubber envisaged. These values are equivalent to a retention of 99.8 % and 99.9 % respectively and correspond to a case with only about 10-20 cm of water above the venturi tube outlets and unfavorable gas dynamic conditions. They can be compared to the design values (DF = 100 and 500, respectively) required to limit ground contamination to the very low level specified by Swedish authorities. 1 fig

R-matrix parameters in reactor applications

International Nuclear Information System (INIS)

Hwang, R.N.

1992-01-01

The key role of the resonance phenomena in reactor applications manifests through the self-shielding effect. The basic issue involves the application of the microscopic cross sections in the macroscopic reactor lattices consisting of many nuclides that exhibit resonance behavior. To preserve the fidelity of such a effect requires the accurate calculations of the cross sections and the neutron flux in great detail. This clearly not possible without viable resonance data. Recently released ENDF/B VI resonance data in the resolved range especially reflect the dramatic improvement in two important areas; namely, the significant extension of the resolved resonance ranges accompanied by the availability of the R-matrix parameters of the Reich-Moore type. Aside from the obvious increase in computing time required for the significantly greater number of resonances, the main concern is the compatibility of the Riech-Moore representation to the existing reactor processing codes which, until now, are based on the traditional cross section formalisms. This purpose of this paper is to summarize our recent efforts to facilitate implementation of the proposed methods into the production codes at ANL

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

Trends on R and D of the innovative nuclear reactors in Japan

International Nuclear Information System (INIS)

Kinoshita, Izumi

2002-01-01

In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

International Nuclear Information System (INIS)

Frajndlich, R.; Sousa, J.A. de.

1985-01-01

A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author) [pt

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)

Experiences from the Swedish programme - heavy water and natural uranium in the Aagesta cogeneration plant

International Nuclear Information System (INIS)

Oestman, Alvar

2002-11-01

A short review of the Swedish programme for nuclear power in the 50's and the 60's is given, and in particular a description of the operating experiences of the Aagesta nuclear cogeneration plant, producing district heating for the south Stockholm area (12 MW el and 68 MW heat ). The original Swedish nuclear programme was built on heavy water and natural uranium and had the objective to construct small nuclear plants in the vicinity of some 10 large cities in south and middle Sweden. Aagesta was the only full-scale plant to be built according to this programme, as Sweden adopted the light-water reactor policy and eventually constructed 12 large reactors at four sites. The report is based on the experiences of the author from his work at the Aagesta plant in the sixties. In an appendix, the experiences from Vattenfall (the Swedish electric utility which took over the operating responsibility for the Aagesta plant), of the plant operation is reviewed

SFCOMPO 2.0 - A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

Science.gov (United States)

Michel-Sendis, Franco; Martinez-González, Jesus; Gauld, Ian

2017-09-01

SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF) samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA). The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Swedish biomass strategies to reduce CO2 emission and oil use in an EU context

International Nuclear Information System (INIS)

Joelsson, Jonas; Gustavsson, Leif

2012-01-01

Swedish energy strategies for transportation, space heating and pulp industries were evaluated with a focus on bioenergy use. The aims were to 1) study trade-offs between reductions in CO 2 emission and oil use and between Swedish reductions and EU reductions, 2) compare the potential contributions of individual reduction measures, 3) quantify the total CO 2 emission and oil use reduction potentials. Swedish energy efficiency measures reduced EU CO 2 emission by 45–59 Mt CO 2 /a, at current biomass use and constant oil use. Doubling Swedish bioenergy use yielded an additional 40 Mt CO 2 /a reduction. Oil use could be reduced, but 36–81 kt of reductions in CO 2 emission would be lost per PJ of oil use reduction. Swedish fossil fuel use within the studied sectors could be nearly eliminated. The expansion of district heating and cogeneration of heat with a high electricity yield were important measures. Plug-in hybrid electric cars reduced CO 2 emission compared with conventional cars, and the difference was larger with increasing oil scarcity. The introduction of black liquor gasification in pulp mills also gave large CO 2 emission reduction. Motor fuel from biomass was found to be a feasible option when coal is the marginal fuel for fossil motor fuel production. -- Highlights: ► Bioenergy is compared to optimized fossil fuel use under different oil availability constraints. ► Swedish strategies are evaluated with respect to CO 2 emission and oil use reduction within Sweden and the EU. ► Efficiency measures give the largest reductions but increased bioenergy use is also important. ► District heating expansion, high electricity yield CHP, increased vehicle efficiency and PHEVs are important options. ► The studied sectors in Sweden could become nearly fossil-fuel free and yield an energy surplus.

Hazard and operability study (Haz Op) of the 2 MW IEA-R1 reactor startup procedures

International Nuclear Information System (INIS)

Sauer, Maria E.L.J.; Correa, Francisco; Sara Neto, Antonio J.; Costa, Carlos A.R. da; Santos, Cilas C. dos; Cardenas, Jose P.N.; Berretta, Jose R.; Neves Conti, Thadeu das

1997-01-01

This work presents the Hazard and Operability Study (Haz Op) applied to startup procedures of the 2 MW IEA-R1 research reactor, at IPEN/CNEN-S P. The Haz Op was developed by reviewing the procedures of the installation startup, in order to identify hazards and/or operational problems caused by deviations in the execution of these routines. This paper summarizes this study. describing some potential problems of relevant importance to safety as well as preventives and/or correctives measures to avoid their occurrence. Besides, an benefits evaluation and the technique limitations is made. (author). 5 refs., 1 tab

Assessment of the effects of neutron fluence on Swedish nuclear pressure vessels

International Nuclear Information System (INIS)

Rao, S.

1980-11-01

Nuclear pressure vessels are subject to neutron irradiation during service causing embrittlement. This is one important factor in the overall problem of reactor vessel integrity. At present the irradiation effects are mainly assessed by the Charpy V-notch test. Two measures of embrittlement are defined: the increase of the ductile/brittle transition temperature and the decrease in the upper-shelf energy. The object of the present work is to assess these changes for the Swedish nuclear pressure vessels. On the basis of data from irradiations carried out in other countries and Swedish surveillance programmes, the expected end of life embrittlement is estimated for Swedish vessels. The results show that the embrittlement of most reactor vessels is expected to be quite small. Oskarshamn 1 and PWR-vessels, however, will probably show moderate changes, the former due to the higher copper content, and the latter due to the high end of life fluences. Some of the vessel materials which exhibit marginal properties in the upper-shelf energy, as measured by the Charpy V-notch impact test, are identified. It is recommended that fracture mechanics analyses be applied in these cases. (author)

Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

CERN Document Server

Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

2000-01-01

Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

Swedish Nuclear Power Inspectorate, Office of Reactor Safety. Research plans for the period 1997-1999

International Nuclear Information System (INIS)

1997-02-01

Office of Reactor Safety research is carried out within the following areas: Safety evaluation, Safety analysis, MTO, Materials and chemistry, Non-Destructive Testing, Strength of materials, Thermohydraulics, Nuclear fuel, Serious accidents and Process control. Research is carried out to fulfill SKIs overall goals in accordance with the directives from the Swedish government and parliament, in particular to be a driving force in safety related work when justified by operating experience, research results and technical progress, towards licensees as well as in international cooperation in safety; to promote the maintenance and development of competence in the safety related work at the SKI as well as the licensees and generally in the country, and as a specific role for the Office of Reactor Safety as designated in the internal routines to take initiative to encourage and carry out research into areas of importance for the Office as well as ensuring that research results are disseminated and used both within SKI and in the general work concerning nuclear safety. Research efforts within the Office of Reactor safety are carried out in the form of separate projects which form part of the priority work plans. Project managers, the necessary personnel resources and the budget for each year are included in the Annual Plan and the work is followed up in the same manner as other efforts. Research is performed in different ways, that can vary from laboratory studies to more consultative efforts, and be organised in many different ways such as examination projects, post-graduate studies, work sponsored at research institutes and companies in Sweden and abroad, collaboration in larger international projects, and participation in conferences which provide an important contribution to keeping SKI personnel informed within their specialist areas

Swedish CO2-emissions 1900-2010: an exploratory note

International Nuclear Information System (INIS)

Kristroem, Bengt; Lundgren, Tommy

2005-01-01

This paper projects Swedish CO 2 -emissions during the period 2000-2010 based on data covering 1900-1999. Swedish climate policy is currently based on the assumption that carbon emissions will increase, ceteris paribus, by 5-15% relative to the 1990 level. This forecast has motivated a number of policy measures, including carbon taxes, subsidies and an 'information package'. We find, however, that CO 2 -emissions may well be lower in the future. This outcome is broadly consistent with the literature on the Environmental Kuznets Curve, which portrays the relationship between emissions and GDP. The key contribution of this paper is that our analysis is based on a long time series. Current literature is invariably based on 'short' panel data sets, while we study a single country through several phases of development. Our analysis also sheds some light on the key importance played by nuclear power for carbon emission projections

Study air ingress into the reactor vessel using ICARE/CATHARE V2.0 in case of severe accident

International Nuclear Information System (INIS)

Gwenaelle Le Dantec; Fichot, F.

2005-01-01

Full text of publication follows: Safety analyses show that core degradation during a severe reactor accident would not be uniform. This was confirmed by TMI2 examinations. In fact, a central region of the core may overheat, melt and flow down to the lower plenum of the reactor while peripheral regions of the core would remain almost intact. Following rupture of the vessel by molten debris, air may be drawn from the containment by natural convection into the reactor coolant system, and react with the intact rods. Studying air ingress into the reactor vessel is of interest because the interaction of air with Zircaloy cladding can strongly affect the evolution of severe accident scenarios. The main effects are heat generation, increasing clad degradation, fission product release and nitriding. In case of air/steam recirculation in the vessel, significant nitriding of cladding can occur. The resulting ZrN phase is characterized by its brittleness and instability under oxidizing conditions, Oxidation of pre-existing ZrN phase layers has been observed to result in violent oxidation and heat release. Therefore, the first consequence for safety is a risk of strong deflagration in the vessel if a large number of rods on which a substantial layer of ZrN has grown are suddenly in contact with oxygen or steam. The second consequence is a late melting of core materials due to the very exothermic oxidation, leading to a late release of materials out of the reactor pressure vessel (RPV). In this paper we present an ICARE/CATHARE V2.0 calculation simulating air ingress into the vessel and in particular to describe the nitriding due to natural convection in the reactor vessel. The basic modeling and the necessary extensions of both ICARE and CATHARE are explained. The natural circulation is calculated to predict the regions of oxygen starvation where nitriding takes place. Key words: air ingress, nitriding, ICARE/CATHARE V2.0. (authors)

Environmental issues associated with R Reactor renovation and startup

International Nuclear Information System (INIS)

Marter, W.L.

1982-01-01

This memorandum identifies the more significant environmental issues that would be associated with renovation and startup of R Reactor to meet future demands for nuclear weapon materials. Some key environmental issues identified are: potential occupational radiation exposures associated with repairing the leaks in the reactor nozzles and thermal shield tanks and with renovating the disassembly basin facilities; the impacts of constructing additional cooling water pumping capacity; the effect of increased pumping on impingement and entrainment at the pump houses; thermal effects on the R Canal-precooler pond-Par Pond ecosystem from increased discharge of cooling water; and effects of increased water flow in Lower Three Runs Creek on biota and on cesium-137 remobilization

The Swedish programme for radioactive waste management

International Nuclear Information System (INIS)

Bjurstroem, S.; Forsstroem, H.

1986-10-01

The following systems and facilities are currently in operation and under implementation: a sea transportation system for all kinds of nuclear waste, a central facility for interim storage of spent fuel (CLAB) and a central underground repository for final disposal of low and medium level reactor waste (SFR). For the remaining steps - final disposal of highly active and longlived radioactive residues - a concept, based on encapsulation of the fuel elements in copper canisters and final storage of the canisters in a repository situated 500 m down in crystalline rock (KBS-3), has been developed and approved by the government in accordance with the Swedish nuclear legislation. Although a feasible method for final disposal of the highly active residues has been shown, the Swedish legislation requires that research be carried out to reach the best possible base for the final decision around the year 2000. In parallel with this a geological investigation programme is carried out to find a suitable site for a final repository. The final site selection is foreseen at the end of the 1990's. All costs for the management of radioactive waste from the nuclear power plants are carried by a fee determined annually. The fee is 0.019 SEK/kWh for 1986

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)

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

Technology and costs for decommissioning the Swedish nuclear power plants

International Nuclear Information System (INIS)

1986-05-01

The study shows that, from the viewpoint of radiological safety, a nuclear power plant can be dismantled immediately after it has been shut down and the fuel has been removed, which is estimated to take about one year. Most of the equipment that will be used in decommissioning is already available and is used routinely in maintenance and rebuilding work at the nuclear power plants. Special equipment need only be developed for dismantlement of the reactor vessel and for demolishing of heavy concrete structures. The dismantling of a nuclear power plant can be accomplished in about five years, with an average labour force of about 200 men. The maximum labour force required for Ringhals 1 has been estimated at about 500 men during the first years, when active systems are being dismantled in a number of fronts in the plant. During the last years when the buildings are being demolished, approximately 50 men are required. In order to limit the labour requirement and the dose burden to the personnel, the material is taken out in as large pieces as possible. The cost of decommissioning a boiling water reactor (BWR) of the size of Ringhals 1 has been estimated to be about MSEK 540 in January 1986 prices, and for a pressurized water reactor (PWR, Ringhals 2) about MSEK 460. The cost for the other Swedish nuclear power plants lie in the range of MSEK 410-760. These are the direct cost for the decommissioning work, to which must be added the costs of transportation and disposal of the decommissioning waste, about 100 000 m/sup3/. These costs have been estimated to be about MSEK 600 for the 12 Swedish reactors. (author)

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

76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

Science.gov (United States)

2011-04-27

... Standard Review Plan, Section 1.0 on Introduction and Interfaces AGENCY: Nuclear Regulatory Commission (NRC... Revision 2 to Standard Review Plan (SRP), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110110573). The Office of New Reactors (NRO...

Magnetite nanoparticles enhance the performance of a combined bioelectrode-UASB reactor for reductive transformation of 2,4-dichloronitrobenzene.

Science.gov (United States)

Wang, Caiqin; Ye, Lu; Jin, Jie; Chen, Hui; Xu, Xiangyang; Zhu, Liang

2017-09-04

Direct interspecies electron transfer (DIET) among the cometabolism microbes plays a key role in the anaerobic degradation of persistent organic pollutants and stability of anaerobic bioreactor. In this study, the COD removal efficiency increased to 99.0% during the start-up stage in the combined bioelectrode-UASB system (R1) with magnetite nanoparticles addition, which was higher than those in the coupled bioelectrode-UASB (R2; 83.2%) and regular UASB (R3; 71.0%). During the stable stage, the increase of 2,4-dichloronitrobenzene (2,4-DClNB) concentration from 25 mg L -1 to 200 mg L -1 did not affect the COD removal efficiencies in R1 and R2, whereas the performance of R3 was deteriorated obviously. Further intermediates analysis indicated that magnetite nanoparticles enhanced the reductive dechlorination of 2,4-DClNB. High-throughput sequencing results showed that the functional microbes like Syntrophobacter and Syntrophomonas which have been reported to favor the DIET, were predominant on the cathode surface of R1 reactor. It is speculated that the addition of magnetite nanoparticles favors the cooperative metabolism of dechlorinating microbes and electricigens during 2,4-DClNB degradation process in the combined bioelectrode-UASB reactor. This study may provide a new strategy to improve the performance of microbial electrolysis cells and enhance the pollutant removal efficiency.

Green light from the Swedish parliament for a renaissance of nuclear energy

International Nuclear Information System (INIS)

Anon.

2010-01-01

On the 17. June 2010, the Swedish parliament passed a law allowing the today's 10 operating reactors to be replaced by new ones at the end of their operational life. These 10 reactors generate half the electrical power of Sweden. The opposition has announced that they will reconsider this law if they win next election. This law will come into effect on the first January 2011. (A.C.)

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

Thermal hydraulic and neutron kinetic coupled simulation of the IPR-R1 Triga reactor

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M. da; Veloso, Maria Auxiliadora F.; Soares, Humbero V., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: betovitor@ig.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq Rede), Rio de Janeiro, RJ (Brazil)

2013-07-01

The nuclear industry and the scientific community have turned the attention for the development of coupled 3D neutron kinetics (NK) and thermal-hydraulic (TH) system codes to investigate specific nuclear reactor transients. Improving in theoretical investigations of complex phenomena in nuclear reactor technology have been increased thanks to numerical methods and computational resources incorporated in nuclear codes. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0 code. The development and the assessment of the thermal-hydraulic RELAP5 code model for the IPR-R1 have been validated for steady state and transient situations and the results were published in preceding works. Results of RELAP5-3D steady state and a transient case presented in this paper show good agreement with experimental data, validating then this model for point kinetic calculations. To supply adequate cross sections to the NK code, the WIMSD5 is being used. First results of steady state calculation using the 3D neutron modeling are being presented in this paper. (author)

Mobile phones, cordless phones and rates of brain tumors in different age groups in the Swedish National Inpatient Register and the Swedish Cancer Register during 1998-2015

OpenAIRE

Hardell, Lennart; Carlberg, Michael

2017-01-01

We used the Swedish Inpatient Register (IPR) to analyze rates of brain tumors of unknown type (D43) during 1998-2015. Average Annual Percentage Change (AAPC) per 100,000 increased with +2.06%, 95% confidence interval (CI) +1.27, +2.86% in both genders combined. A joinpoint was found in 2007 with Annual Percentage Change (APC) 1998-2007 of +0.16%, 95% CI -0.94, +1.28%, and 2007-2015 of +4.24%, 95% CI +2.87, +5.63%. Highest AAPC was found in the age group 20-39 years. In the Swedish Cancer Regi...

k{sub 0}-INAA method at the pneumatic station of the IEA-R1 nuclear research reactor. Application to geological samples

Energy Technology Data Exchange (ETDEWEB)

Mariano, Davi B.; Figueiredo, Ana Maria G.; Semmler, Renato, E-mail: davimariano@usp.br, E-mail: anamaria@ipen.br, E-mail: rsemmler@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

There is a significant number of analytically important elements, when geological samples are concerned, whose activation products are short-lived (seconds to minutes) or medium-lived radioisotopes (minutes to hours). As part of the process of implementation of the k{sub 0}-INAA standardization method at the Neutron Activation Laboratory (LAN-IPEN), Sao Paulo, Brazil, this study presents the results obtained for the analysis of short and medium-lived nuclides in geological samples by k{sub 0}-INAA using the program k{sub 0}-IAEA, provided by The International Atomic Energy Agency (IAEA). The elements Al, Dy, Eu, Na, K, Mn, Mg, Sr, V and Ti were determined with respect to gold ({sup 197}Au) using the pneumatic station facility of the IEA-R1 5 MW swimming pool nuclear research reactor, Sao Paulo. Characterization of the pneumatic station was carried out by using the -bare triple-monitor- method with {sup 197}Au-{sup 96}Zr-{sup 94}Zr. The Certified Reference Material IRMM-530R Al-0,1% Au alloy, high purity zirconium, Ni and Lu comparators were irradiated. The efficiency curves of the gamma-ray spectrometer used were determined by measuring calibrated radioactive sources at the usually utilized counting geometries. The method was validated by analyzing the reference materials basalt BE-N (IWG-GIT), basalt JB- 1 (GSJ), andesite AGV-1 (USGS), granite GS-N (IWG-GIT), SOIL-7 (IAEA) and sediment Buffalo River Sediment (NIST-BRS-8704), which represent different geological matrices. The concentration results obtained agreed with certified, reference and recommended values, showing relative errors less than 10% for most elements. (author)

An ocean of stress? The relationship between psychosocial workload and mental strain among engine officers in the Swedish merchant fleet.

Science.gov (United States)

Rydstedt, Leif W; Lundh, Monica

2010-01-01

The first purpose of this study was to compare the psychosocial working conditions and mental health of our sample of maritime engine officers with a sample of British shore-based professional engineers. The second purpose was to analyse the relationship between the psychosocial working conditions onboard and mental strain for the Swedish maritime engine officers. There were a total of 731 engine officers in the Swedish merchant fleet, almost all males with higher education. The British comparison sample consisted of 312 professional shore-based engineers. A questionnaire was distributed to the Swedish engine officers with a modified version of the JCQ for the DC-S model, the Role conflict and Ambiguity scale, and two items on family-work inter-role conflicts (WFI/FWI), as workload indicators. The General Health Questionnaire (GHQ12) and Perceived Stress Scale (PSS10) were used as strain indicators. There were no significant differences in perceived job stain or in WFI/FWI between the Swedish engine officers and the British professional engineers in perceived job strain. While the British shore-based engineers reported significantly higher role ambiguity the Swedish engine officers perceived a significantly higher degree of role conflict and higher perceived stress. Hierarchic linear regression analysis showed that the Role Stress was strongly related to perceived stress (R(2) = 0.319) as well as to mental health (R(2) = 0.222). When introduced in the second step the DC-S model was significantly related to the outcome measures, as was WFI/FWI when finally introduced. The main source of the high degree of perceived stress among the engine officers does not seem to be the job content but may rather be understood from an interactional perspective, where conflicting requirements are directed towards the individual officer. It can be assumed that the fast technological and organizational changes and the increased pressure for economic profitability that characterize the

Crystal structure of (1S,3R,8R,9R-2,2-dichloro-3,7,7-trimethyl-10-methylenetricyclo[6.4.0.01,3]dodecan-9-ol

Directory of Open Access Journals (Sweden)

Ahmed Benzalim

2016-08-01

Full Text Available The title compound, C16H24Cl2O, was synthesized by treating (1S,3R,8S,9R,10S-2,2-dichloro-3,7,7,10-tetramethyl-9,10-epoxytricyclo[6.4.0.01,3]dodecane with a concentrated solution of hydrobromic acid. It is built up from three fused rings: a cycloheptane ring, a cyclohexyl ring bearing alkene and hydroxy substituents, and a cyclopropane ring bearing two chlorine atoms. The asymmetric unit contains two molecules linked by an O—H...O hydrogen bond. In the crystal, further O—H...O hydrogen bonds build up an R44(8 cyclic tetramer. One of the molecules presents disorder that affects the seven-membered ring. In both molecules, the six-membered rings display a chair conformation, whereas the seven-membered rings display conformations intermediate between boat and twist-boat for the non-disordered molecule and either a chair or boat and twist-boat for the disordered molecule owing to the disorder. The absolute configuration for both molecules is 1S,3R,8R,9R and was deduced from the chemical pathway and further confirmed by the X-ray structural analysis.

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

Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

International Nuclear Information System (INIS)

Bergdahl, B.G.

1998-05-01

The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

Swedish subseabed store - phase 1 nears completion

International Nuclear Information System (INIS)

Daglish, James

1987-01-01

The paper concerns the storage of radioactive waste in the subseabed in Sweden. The wastes are low- and intermediate-level reactor wastes arising from the Swedish nuclear power programme. The repository is a cavern which has been excavated under the seabed in the Baltic Sea, about a kilometre out from shore. The specifications of the repository are given, along with the volume of the radioactive wastes to be stored in it. (UK)

Management of European fast reactor R and D

International Nuclear Information System (INIS)

Judd, A.M.; Sheriff, N.

1993-01-01

Since 1984 government-funded fast reactor R and D in France, Germany and the UK has been run as a collaborative activity, and since 1988 as a unified programme in support of the design and construction of the advanced European Fast Reactor. This paper describes the international management structure which has been set up, and the means used to control the work. It is written from the point of view of those engaged in the project, and makes no attempt at a formal analysis of the structure. The main difficulty is that control of funding remains with the three governments. The R and D programme has to be managed so that it meets the needs of each government separately as well as the designers' requirements. To start with the management structure was excessively bureaucratic, but it has become more flexible and efficient. This has happened as the initial nationalistic suspicions have broken down, and the staff engaged in the work have learnt more about each others' ways of working so that an atmosphere of trust and inter-dependence has grown up. (This paper was written before the changes in UK policy on fast reactor development were announced in November 1992). (Author)

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

The development of fast reactors - Effects on the Swedish system of management of spent fuel; Utveckling av snabba reaktorer - Paaverkan paa det svenska systemet foer hantering av anvaent braensle

Energy Technology Data Exchange (ETDEWEB)

Hans Forsstroem, Hans [SKB International AB, Stockholm (Sweden)

2013-09-15

. In this context it should be considered that fast reactors will generate their own plutonium, as breeder reactors. Plutonium from other reactors will thus only be needed for the first years of operation. To provide a basis for the answer to the question if the Swedish spent fuel is a resource or a waste this report provides an overview of the present development status for fast reactors and their potential for large scale commercial use. It further describes the impact on the Swedish system for management of spent nuclear fuel if the fuel were to be reprocessed and the uranium and plutonium reused as fuel for fast reactors or for the present reactors.

Fast reactors: R and D targets and outlook for their introduction

International Nuclear Information System (INIS)

Poplavsky, V.; Barre, B.; Aizawa, K.

1997-01-01

In this paper the current status of fast reactors development is briefly outlined, including experimental, demonstration, and commercial installations. Data on the experience gained in development and operation of NPPs with reactors of this type are presented. The issues are discussed in connection with possibilities of fast reactor development in the nuclear power structure for the near (up to 2010-2020) and distant future. In the final part of the paper, an analysis is given of possible ways for R and D development in the field of NPPs with fast neutron reactors. (author)

WWER safety investigations on LR-0 reactor

International Nuclear Information System (INIS)

Mikus, J.

2001-01-01

A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

Operating experience from Swedish nuclear power plants, 1999

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

The total generation of electricity from Swedish nuclear power plants was 70.1 TWh during 1999, which is slightly more than the mean value for the last five years. The total electricity consumption decreased by one percent, compared with 1998, to a total of 142.3 TWh, due to an unusually warm summer and autumn. The abundant supply of hydroelectric power resulted in comparatively extensive load-following operation by the nuclear plants during the year. Production losses due to low demand totalled 3.0 TWh. The closure of Barsebaeck 1 will result in a capacity reduction exceeding 4 TWh per year. The hydroelectric power production was 70 TWh, which was 6 TWh more than during a normal year, i.e. a year with average rainfall. The remaining production sources, mainly from solid fuel plants combined with district heating contributed 9 TWh. Electricity generation by means of wind power is still increasing. There are now about 470 wind power stations, which produced 0.3 TWh during the year. The total electricity generation totalled 149.8 TWh, a three percent decrease compared with 1998. The preliminary figures for export were 15.9 TWh and for import 8.4 TWh. The figures above are calculated from the preliminary production result. A comprehensive report on electric power supply and consumption in Sweden is provided in the 1999 Annual Report from the Swedish Power Association. The unit capability factor for the PWRs at Ringhals averaged 91%, while the BWRs averaged 82% mainly due to the extended outages. The BWR reactors at Forsmark averaged as much as 93%. Forsmark 1 experienced the shortest refuelling outage ever in Sweden, only 9 days and 20 hours. In May, Oskarshamn 2 passed a historical milestone - the unit produced 100 TWh since connection to the grid in 1974. The final production day for Barsebaeck 1, which had been in commercial operation since 1975, was on November 30 when a decision by the Swedish Government revoked the operating licence. Three safety-related events

SFCOMPO 2.0 – A relational database of spent fuel isotopic measurements, reactor operational histories, and design data

Directory of Open Access Journals (Sweden)

Michel-Sendis Franco

2017-01-01

Full Text Available SFCOMPO-2.0 is a database of experimental isotopic concentrations measured in destructive radiochemical analysis of spent nuclear fuel (SNF samples. The database includes corresponding design description of the fuel rods and assemblies, relevant operating conditions and characteristics of the host reactors necessary for modelling and simulation. Aimed at establishing a thorough, reliable, and publicly available resource for code and data validation of safety-related applications, SFCOMPO-2.0 is developed and maintained by the OECD Nuclear Energy Agency (NEA. The SFCOMPO-2.0 database is a Java application which is downloadable from the NEA website.

Application of the k0-NAA method at the HANARO research reactor

International Nuclear Information System (INIS)

Jong-Hwa Moon; Sun-Ha Kim; Yong-Sam Chung; Young-Jin Kim

2007-01-01

The k 0 -standardization method (k 0 -NAA) is known as one of the most remarkable progresses of the NAA with its many advantages. For the application of k 0 -NAA method at the NAA 1 irradiation position where the neutrons are well thermalized in the HANARO research reactor, KAERI, Korea, the determination of the reactor neutron spectrum parameters such as α and f have been carried out. The measured values of a and f using the 'Cd-ratio' triple monitor method were 0.127±0.022 and 1010±70, respectively. To evaluate the applicability of k 0 -NAA in our analytical system, the analysis of three kinds of SRMs was executed. The analytical results showed that the relative error of most of the elements was less than 10% and the U-scores were within 2. It is turned out that the procedure of the k 0 -NAA in the HANARO research reactor is available for a practical application in the environmental fields. (author)

High field induced magnetic transitions in the Y0.7E r0.3F e2D4.2 deuteride

Science.gov (United States)

Paul-Boncour, V.; Guillot, M.; Isnard, O.; Hoser, A.

2017-09-01

The influence of the partial Er for Y substitution on the crystal structure and magnetic properties of YF e2D4.2 has been investigated by high field magnetization and neutron diffraction experiments. Y0.7E r0.3F e2D4.2 compound crystallizes in the same monoclinic structure as YF e2D4.2 described in P c (P1c1) space group with D atoms located in 18 different tetrahedral interstitial sites. A cell volume contraction of 0.6% is observed upon Er substitution, inducing large modification of the magnetic properties. Electronic effect of D insertion as well as lowering of crystal symmetry are important factors determining the magnetic properties of Fe sublattice, which evolves towards more delocalized behavior and modifying the Er-Fe exchange interactions. In the ground state, the Er and Fe moments are arranged ferrimagnetically within the plane perpendicular to the monoclinic b axis and with average moments mEr=6.4 (3 ) μBEr-1 and mFe=2.0 (1 ) μBFe-1 at 10 K. Upon heating, mEr decreases progressively until TEr=55 K . Between 55 K and 75 K, the Fe sublattice undergoes a first-order ferromagnetic-antiferromagnetic (FM-AFM) transition with a cell volume contraction due to the itinerant metamagnetic behavior of one Fe site. In the AFM structure, mFe decreases until the Néel temperature TN=125 K . At high field, two different types of field induced transitions are observed. The Er moments become parallel to the Fe one and saturates to the E r3 + free ion value, leading to an unusual field induced FM arrangement at a transition field BTrans of only 78 kG below 30 K. Then above TM0=66 K , an AFM-FM transition of the Fe sublattice, accompanied by a cell volume increase is observed. BTrans increases linearly versus temperature and with a larger d BTrans/d T slope than for YF e2D4.2 . This has been explained by the additional contribution of Er induced moments above BTrans.

Swedish experiences in implementing national and international safeguards

International Nuclear Information System (INIS)

Nilsson, A.; Elborn, M.; Grahn, P.

1991-01-01

This paper reports that international safeguards have been applied in Sweden since the early 70s. Experiences have been achieved from exclusive bilateral and trilateral control followed by NPT safeguards in 1975. The Swedish State System for accountancy and Control (SSAC) includes all regulations that follows from prevailing obligations regarding the peaceful uses of nuclear material. The system has been developed in cooperation between the national authority, the Swedish Nuclear Power Inspectorate (SKI) and the Swedish nuclear industry. The paper presents experiences from the practical implementation of the SSAC and the IAEA safeguards system, gained by the SKI and the nuclear industry, respectively. Joint approaches and solutions to some significant safeguards issues are presented. The cooperation between the nuclear industry and the authority in R and D activities, in particular with respect to the Swedish Support Program is highlighted, e.g. the use of nuclear facilities in development or training tasks. some of the difficulties encountered with the system are also touched upon

Genetic anticipation in Swedish Lynch syndrome families.

Directory of Open Access Journals (Sweden)

Jenny von Salomé

2017-10-01

Full Text Available Among hereditary colorectal cancer predisposing syndromes, Lynch syndrome (LS caused by mutations in DNA mismatch repair genes MLH1, MSH2, MSH6 or PMS2 is the most common. Patients with LS have an increased risk of early onset colon and endometrial cancer, but also other tumors that generally have an earlier onset compared to the general population. However, age at first primary cancer varies within families and genetic anticipation, i.e. decreasing age at onset in successive generations, has been suggested in LS. Anticipation is a well-known phenomenon in e.g neurodegenerative diseases and several reports have studied anticipation in heritable cancer. The purpose of this study is to determine whether anticipation can be shown in a nationwide cohort of Swedish LS families referred to the regional departments of clinical genetics in Lund, Stockholm, Linköping, Uppsala and Umeå between the years 1990-2013. We analyzed a homogenous group of mutation carriers, utilizing information from both affected and non-affected family members. In total, 239 families with a mismatch repair gene mutation (96 MLH1 families, 90 MSH2 families including one family with an EPCAM-MSH2 deletion, 39 MSH6 families, 12 PMS2 families, and 2 MLH1+PMS2 families comprising 1028 at-risk carriers were identified among the Swedish LS families, of which 1003 mutation carriers had available follow-up information and could be included in the study. Using a normal random effects model (NREM we estimate a 2.1 year decrease in age of diagnosis per generation. An alternative analysis using a mixed-effects Cox proportional hazards model (COX-R estimates a hazard ratio of exp(0.171, or about 1.19, for age of diagnosis between consecutive generations. LS-associated gene-specific anticipation effects are evident for MSH2 (2.6 years/generation for NREM and hazard ratio of 1.33 for COX-R and PMS2 (7.3 years/generation and hazard ratio of 1.86. The estimated anticipation effects for MLH1

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

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

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

R and D on fast reactor fuel reprocessing

International Nuclear Information System (INIS)

Subba Rao, R.V.; Vijaya Kumar, V.; Natarajan, R.

2012-01-01

Development of Fast Reactor Fuel Reprocessing technology, with low out of pile inventory, is carried out at the Indira Gandhi Centre for Atomic Research (IGCAR). Based on the successful R and D programme which addressed specific issues of fast reactor fuels, a pilot plant called CORAL was set up. This plant is operational since 2003 and several reprocessing campaigns with spent FBTR fuels of varying burnups have been carried out. Based on the valuable operating experience of CORAL, the design of demonstration fast reactor fuel reprocessing plant (DFRP) and the commercial reprocessing plant, FRP have been taken up. Concurrently R and D efforts are continuing for improving the process and equipment performance apart from reducing the waste volumes and the radiation exposures to the operating personnel. Some important R and D efforts are highlighted in the paper. Reducing the dissolution time is one of the vital area of investigation especially for the high plutonium bearing MOX fuels which are known to dissolve slowly. To address this as well as criticality issues, continuous dissolvers are being developed. Solvent extraction based process is employed for getting highly pure nuclear grade uranium and plutonium. In view of the lower cooling time the fission product activity in the spent fuel is higher, formulation of process flowsheet with reduced number of solvent extraction cycles to improve the decontamination of ruthenium and zirconium without the formation of second organic phase due to plutonium loading, is under investigation. Retention of plutonium in lean organic is another issue to be addressed as otherwise it would lead to further deterioration of the solvent on storage. Several reagents to effectively wash the lean solvent have been investigated and flowsheets have been formulated to recover the retained plutonium with minimum secondary wastes. Partitioning of uranium and plutonium is an important step and methods reported in the literature have inherent

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)

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)

The Swedish Version of the Ritvo Autism and Asperger Diagnostic Scale: Revised (RAADS-R). A Validation Study of a Rating Scale for Adults

Science.gov (United States)

Andersen, Lisa M. J.; Naswall, Katharina; Manouilenko, Irina; Nylander, Lena; Edgar, Johan; Ritvo, Riva Ariella; Ritvo, Edward; Bejerot, Susanne

2011-01-01

There is a paucity of diagnostic instruments for adults with autism spectrum disorder (ASD). This study evaluates the psychometric properties of the Swedish version of the Ritvo Autism and Asperger Diagnostic Scale-Revised (RAADS-R), an 80-item self-rating scale designed to assist clinicians diagnosing ASD in adults. It was administered to 75…

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

ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: The European Reactor Analysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R and D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core

Final report on the Swedish participation in PISC II

International Nuclear Information System (INIS)

Hoegberg, K.; Zettewall, T.

1986-08-01

The aim of the project is to evaluate the reliability of test methods for reactor pressure vessels and to identify the proper methods for defect control and to inform about the results. Four test plates planted defect have been investigated by 50 testing teams from 13 countries. Swedish testing has shown acceptable data for the detection of defects when using high sensitivity tests. (G.B.)

Swedish biomass strategies to reduce CO{sub 2} emission and oil use in an EU context

Energy Technology Data Exchange (ETDEWEB)

Joelsson, Jonas [Ecotechnology and Environmental Science, Mid Sweden University, SE-831 25 Oestersund (Sweden); Gustavsson, Leif [Linnaeus University, SE-351 95 Vaexjoe (Sweden)

2012-07-15

Swedish energy strategies for transportation, space heating and pulp industries were evaluated with a focus on bioenergy use. The aims were to 1) study trade-offs between reductions in CO{sub 2} emission and oil use and between Swedish reductions and EU reductions, 2) compare the potential contributions of individual reduction measures, 3) quantify the total CO{sub 2} emission and oil use reduction potentials. Swedish energy efficiency measures reduced EU CO{sub 2} emission by 45-59 Mt CO{sub 2}/a, at current biomass use and constant oil use. Doubling Swedish bioenergy use yielded an additional 40 Mt CO{sub 2}/a reduction. Oil use could be reduced, but 36-81 kt of reductions in CO{sub 2} emission would be lost per PJ of oil use reduction. Swedish fossil fuel use within the studied sectors could be nearly eliminated. The expansion of district heating and cogeneration of heat with a high electricity yield were important measures. Plug-in hybrid electric cars reduced CO{sub 2} emission compared with conventional cars, and the difference was larger with increasing oil scarcity. The introduction of black liquor gasification in pulp mills also gave large CO{sub 2} emission reduction. Motor fuel from biomass was found to be a feasible option when coal is the marginal fuel for fossil motor fuel production. -- Highlights: Black-Right-Pointing-Pointer Bioenergy is compared to optimized fossil fuel use under different oil availability constraints. Black-Right-Pointing-Pointer Swedish strategies are evaluated with respect to CO{sub 2} emission and oil use reduction within Sweden and the EU. Black-Right-Pointing-Pointer Efficiency measures give the largest reductions but increased bioenergy use is also important. Black-Right-Pointing-Pointer District heating expansion, high electricity yield CHP, increased vehicle efficiency and PHEVs are important options. Black-Right-Pointing-Pointer The studied sectors in Sweden could become nearly fossil-fuel free and yield an energy

Dynamic modelling of nitrous oxide emissions from three Swedish sludge liquor treatment systems

DEFF Research Database (Denmark)

Lindblom, E.; Arnell, M.; Flores-Alsina, X.

2014-01-01

The objective of this paper is to model the dynamics and validate the results of nitrous oxide (N2O)emissions from three Swedish nitrifying/denitrifying, nitritation and anammox systems treating real anaerobic digester sludge liquor. The Activated Sludge Model No. 1 is extended to describe N2O...... production by both heterotrophic and autotrophic denitrification. In addition, mass transfer equations are implemented to characterize the dynamics of N2O in the water and the gas phases.The biochemical model is simulated and validated for two hydraulic patterns: (1) a sequencing batch reactor; and, (2...

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

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

Operating experience from Swedish nuclear power plants

International Nuclear Information System (INIS)

1998-01-01

During 1997 the PWRs in Ringhals performed extremely well (capability factors 85-90%), the unit Ringhals 2 reached the best capability factor since commercial operation started in 1976. The BWRs made an average 76% capability, which is somewhat less than in 1996. The slightly reduced capability derives from ongoing modernization projects at several units. At the youngest plants, Forsmark 3 and Oskarshamn 3, capability and utilization were very high. Events and data for 1997 are given for each reactor, together with operational statistics for the years 1990-1997. A number of safety-related events are reported, which occurred st the Swedish plants during 1997. These events are classified as level 1 or higher on the international nuclear event scale (INES)

Operating experience from Swedish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-06-01

During 1997 the PWRs in Ringhals performed extremely well (capability factors 85-90%), the unit Ringhals 2 reached the best capability factor since commercial operation started in 1976. The BWRs made an average 76% capability, which is somewhat less than in 1996. The slightly reduced capability derives from ongoing modernization projects at several units. At the youngest plants, Forsmark 3 and Oskarshamn 3, capability and utilization were very high. Events and data for 1997 are given for each reactor, together with operational statistics for the years 1990-1997. A number of safety-related events are reported, which occurred st the Swedish plants during 1997. These events are classified as level 1 or higher on the international nuclear event scale (INES).

Design and properties of marine reactors and associated R and D

Energy Technology Data Exchange (ETDEWEB)

Gagarinski, A; Ignatiev, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

Design and properties of marine reactors and associated R and D

International Nuclear Information System (INIS)

Gagarinski, A.; Ignatiev, V.; Devell, L.

1996-05-01

The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs

Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

International Nuclear Information System (INIS)

Hadek, J.; Grundmann, U.

1989-09-01

Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

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)

Swedish CO{sub 2}-emissions 1900-2010: an exploratory note

Energy Technology Data Exchange (ETDEWEB)

Kristroem, Bengt; Lundgren, Tommy [Swedish Univ. of Agricultural Sciences, Dept. of Forest Economics, Umeaa (Sweden)

2005-06-01

This paper projects Swedish CO{sub 2}-emissions during the period 2000-2010 based on data covering 1900-1999. Swedish climate policy is currently based on the assumption that carbon emissions will increase, ceteris paribus, by 5-15% relative to the 1990 level. This forecast has motivated a number of policy measures, including carbon taxes, subsidies and an 'information package'. We find, however, that CO{sub 2}-emissions may well be lower in the future. This outcome is broadly consistent with the literature on the Environmental Kuznets Curve, which portrays the relationship between emissions and GDP. The key contribution of this paper is that our analysis is based on a long time series. Current literature is invariably based on 'short' panel data sets, while we study a single country through several phases of development. Our analysis also sheds some light on the key importance played by nuclear power for carbon emission projections. (Author)

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

Energy Technology Data Exchange (ETDEWEB)

Hellstrand, E; Andersson, T L; Brunfelter, B; Kockum, J; Londen, S O; Tiren, L I

1965-12-15

In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

Fuel performance in the Barsebeck boiling water reactors (Unit 1 and 2)

International Nuclear Information System (INIS)

Norman, B.

1979-01-01

Sydkraft is the largest privately owned utility in Sweden. It serves about 20% of the Swedish population with about 12 TWh of electric power per year, of which 64% is nuclear (1978 figures). The two identical 590 MWE ASEA-ATOM boiling water reactors in Barsebeck have been in operation since 1975 and 1977 respectively. Fission product activity in the primary circuits and in the off-gas systems is extremely low and indicate a near perfect fuel condition. Operating restrictions limiting the effect of pellet cladding interaction have been in use since initial start-up and testing. A few events involving rapid power increases above the preconditioned power level have occurred without causing fuel failures. It is believed that an analysis of power reactor operational transients, which did not cause fuel failures, can be useful to design more adequate and less conservative rules for the operation of nuclear reactor cores

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)

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

Nuclide content in reactor waste

International Nuclear Information System (INIS)

1981-11-01

Certain corrosion and fission products of importance in reactor waste management cannot be measured by gammaspectrometric techniques. In this study, a method is suggested by which the occurence of such nuclides can be quantitatively related to suitable gamma-emitters of similar origin. The method is tested by statistical analysis on the waste data recorded from two Swedish nuclear power plants. As this method is not applicable for Carbon-14, this nuclide was measured directly in spent ion exchange resins from three Finnish and Swedish power plants. (author)

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)

Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

International Nuclear Information System (INIS)

Gustavsson, Veine

2002-11-01

The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

Final report on the IAEA research contracts No. 1194/RB, 1194/R1/RB and 1194/R2/RB

International Nuclear Information System (INIS)

Zobor, E.; Janosy, J.S.; Szentgali, A.

1980-09-01

The final report summarizes the research activities made in the framework of the IAEA Research Contracts No. 1194/RB, 1194/R1/RB and 1194/R2/RB. A multilevel hierarchical control system is treated which uses weakly-coupled low dimensional subsystems under the supervision of a dynamic coordinator program. This self-organizing adaptive control system was checked by a 5 MW research reactor. As an example the paper describes the experimental computer control system of the 5 MW WWR-SM research reactor, where the reactor power and outlet temperature have been controlled on the basis of the treated control concept since 1978. (author)

Industrial electricity demand and energy efficiency policy: The role of price changes and private R and D in the Swedish pulp and paper industry

International Nuclear Information System (INIS)

Henriksson, Eva; Söderholm, Patrik; Wårell, Linda

2012-01-01

The objective of this paper is to analyze electricity demand behaviour in the Swedish pulp and paper industry in the context of the increased interest in so-called voluntary energy efficiency programs. In these programs tax exemptions are granted if the participating firms carry out energy efficiency measures following an energy audit. We employ a panel data set of 19 pulp and paper firms, and estimate both the own- and cross-price elasticities of electricity demand as well as the impact of knowledge accumulation following private R and D on electricity use. The empirical results show that electricity use in the Swedish pulp and paper industry is relatively own-price insensitive, and the self-reported electricity savings following the voluntary so-called PFE program support the notion of important information asymmetries at the company level. However, the results display that already in a baseline setting pulp and paper firms tend to invest in private R and D that have electricity saving impacts, and our model simulations suggest that up to about one-third of the industry sector's self-reported electricity savings in PFE could be attributable to pure baseline effects. Future evaluations of voluntary energy efficiency programs must increasingly recognize the already existing incentives to reduce energy use in energy-intensive industries. - Highlights: ► We analyze electricity demand behaviour in the Swedish pulp and paper industry. ► An important context is the voluntary energy efficiency programs PFE. ► The electricity savings following PFE are significant, but price responses are low. ► Still, already in a baseline setting firms tend to invest in electricity-saving R and D. ► These baseline issues are not adequately addressed in PFE.

MethylMix 2.0: an R package for identifying DNA methylation genes.

Science.gov (United States)

Cedoz, Pierre-Louis; Prunello, Marcos; Brennan, Kevin; Gevaert, Olivier

2018-04-14

DNA methylation is an important mechanism regulating gene transcription, and its role in carcinogenesis has been extensively studied. Hyper and hypomethylation of genes is a major mechanism of gene expression deregulation in a wide range of diseases. At the same time, high-throughput DNA methylation assays have been developed generating vast amounts of genome wide DNA methylation measurements. We developed MethylMix, an algorithm implemented in R to identify disease specific hyper and hypomethylated genes. Here we present a new version of MethylMix that automates the construction of DNA-methylation and gene expression datasets from The Cancer Genome Atlas (TCGA). More precisely, MethylMix 2.0 incorporates two major updates: the automated downloading of DNA methylation and gene expression datasets from TCGA and the automated preprocessing of such datasets: value imputation, batch correction and CpG sites clustering within each gene. The resulting datasets can subsequently be analyzed with MethylMix to identify transcriptionally predictive methylation states. We show that the Differential Methylation Values created by MethylMix can be used for cancer subtyping. olivier.gevaert@stanford.edu. https://bioconductor.org/packages/release/bioc/manuals/MethylMix/man/MethylMix.pdf. MethylMix 2.0 was implemented as an R package and is available in bioconductor.

Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

International Nuclear Information System (INIS)

Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

2009-01-01

R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

Removal of FePO4 and Fe3(PO4)2 crystals on the surface of passive fillers in Fe0/GAC reactor using the acclimated bacteria

International Nuclear Information System (INIS)

Lai, Bo; Zhou, Yuexi; Yang, Ping; Wang, Juling; Yang, Jinghui; Li, Huiqiang

2012-01-01

Highlights: ► Fe 3 (PO 4 ) 2 and FePO 4 crystals would weaken treatment efficiency of Fe 0 /GAC reactor. ► Fe 3 (PO 4 ) 2 and FePO 4 crystals could be removed by the acclimated bacteria. ► FeS and sulfur in the passive film would be removed by the sulfur-oxidizing bacteria. ► Develop a cost-effective bio-regeneration technology for the passive fillers. - Abstract: As past studies presented, there is obvious defect that the fillers in the Fe 0 /GAC reactor begin to be passive after about 60 d continuous running, although the complicated, toxic and refractory ABS resin wastewater can be pretreated efficiently by the Fe 0 /GAC reactor. During the process, the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film are formed by the reaction between PO 4 3− and Fe 2+ /Fe 3+ . Meanwhile, they obstruct the formation of macroscopic galvanic cells between Fe 0 and GAC, which will lower the wastewater treatment efficiency of Fe 0 /GAC reactor. In this study, in order to remove the Fe 3 (PO 4 ) 2 and FePO 4 crystals on the surface of the passive fillers, the bacteria were acclimated in the passive Fe 0 /GAC reactor. According to the results, it can be concluded that the Fe 3 (PO 4 ) 2 and FePO 4 crystals with high density in the passive film could be decomposed or removed by the joint action between the typical propionic acid type fermentation bacteria and sulfate reducing bacteria (SRB), whereas the PO 4 3− ions from the decomposition of the Fe 3 (PO 4 ) 2 and FePO 4 crystals were released into aqueous solution which would be discharged from the passive Fe 0 /GAC reactor. Furthermore, the remained FeS and sulfur (S) in the passive film also can be decomposed or removed easily by the oxidation of the sulfur-oxidizing bacteria. This study provides some theoretical references for the further study of a cost-effective bio-regeneration technology to solve the passive problems of the fillers in the zero-valent iron (ZVI) or Fe 0 /GAC reactor.

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

Magneto-resistance of Si0,97Ge0,03 whiskers irradiated by reactor fast neutrons

International Nuclear Information System (INIS)

Pavlovska, N.T.; Litovchenko, P.G.; Karpenko, A.Ya.; Uhryn, Yu.O.; Pavlovskyj, Yu.V.; Ostrovskii, I.P.; Khoverko, Yu.M.

2012-01-01

The influence of reactor fast-neutrons irradiation by the fluence of 8,6·10 17 n/cm 2 and strong magnetic field (up to 14 T) on resistance of Si 1-x Ge x (x = 0.03) whiskers in the temperature range of 4,2 - 300 K is studied. The activation energy of the of impurity levels is calculated. The interpretation of changes in the magneto-resistance is proposed

The Canadian R and D program targeted at CANDU reactors

International Nuclear Information System (INIS)

Moeck, E.O.

1988-01-01

CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

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)

Qualification of JEFF3.1.1 library for high conversion reactor calculations using the ERASME/R experiment

Energy Technology Data Exchange (ETDEWEB)

Vidal, J. F.; Noguere, G.; Peneliau, Y.; Santamarina, A. [CEA, DEN, DER/SPRC/LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance (France)

2012-07-01

With its low CO{sub 2} production, Nuclear Energy appears to be an efficient solution to the global warming due to green-house effect. However, current LWR reactors are poor uranium users and, pending the development of Fast Neutron Reactors, alternative concepts of PWR with higher conversion ratio (HCPWR) are being studied again at CEA, first studies dating from the middle 80's. In these French designs, low moderation ratio has been performed by tightening the lattice pitch, achieving a conversion ratio of 0.8-0.9 with a MOX fuel coming from PWR UOX recycling. Theses HCPWRs are characterized by a harder neutron spectrum and the calculation uncertainties on the fundamental neutronics parameters are increased by a factor 3 regarding a standard PWR lattice, due to the major contribution of the Plutonium isotopes and of the epithermal energy range to the reaction rates. In order to reduce these uncertainties, a 3-year experimental validation program called ERASME has been performed by CEA from 1984 to 1986 in the EOLE reactor. Monte Carlo analysis of the ERASME/R experiments with the Monte Carlo code TRIPOLI4 allowed the qualification of the recommended JEFF.3.1.1 library for major neutronics parameters. K{sub eff} of the MOX under-moderated lattice is over-predicted by 440 {+-} 830 pcm (2{sigma}); the conversion ratio, indicator of the good use of uranium, is also slightly over-predicted: 2 % {+-} 4 % (2{sigma}) and the same for B4C absorber rods worth and soluble boron worth, over-predicted by 2 %, both in the 2 standard deviations range. The radial fission maps of heterogeneities (water-holes, B4C and fertile rods) are well reproduced: maximal (C-E)/E dispersion is 1.3 %, maximal power peak error is 2.7 %. The void reactivity worth is the only parameter poorly calculated with an overprediction of +12.4% {+-} 1.5%. ERASME/R analysis of MOX reactivity, void effect and spectral indexes will contribute to the reevaluation of {sup 241}Am and Plutonium isotopes

Report on the contacts between Swedish and Danish authorities at the construction of the Barsebaeck NPP

International Nuclear Information System (INIS)

Nilsson, Tore

2002-04-01

The contacts between Danish and Swedish authorities before the building of the Barsebaeck nuclear power plants are reviewed. The information exchange started in 1968 (operation of the first reactor started in 1975) and the general optimistic view on nuclear power at the time was reflected on the positive view on the reactor installations from the Danish side. During the 1970s, the attitudes changed and a lot of efforts were made from the Danish authorities to analyze the safety aspects of the reactors

Thermal instability observations during ramp tests in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Roennberg, G.; Kjaer-Pedersen, N.

1984-01-01

A series of ramp tests on ENC-built BWR fuel from the Big Rock Point reactor was performed in September 1982 in the Studsvik R2 Reactor. The tests involved segmented rods with a burnup of 18 MWd/KgU, and constituted part of the Fuel Performance Improvement Program sponsored by the United States Department of Energy. Rods of different designs were tested. The reference design had solid, dished pellets and was unpressurized. The alternative designs were annular pellets and sphere-pac. Some of the rods with annular pellets were prepressurized, and some were not. During the ramp tests the rod power is controlled by a helium depressurization loop which causes a strictly linear power ramp versus time. The thermal output of the test rig is measured calorimetrically, the data immediately being recorded on a strip chart and later processed by a computer. Furthermore, elongation detectors permit the immediate recording of the rod length variation versus time. For some of the rods the thermal output went constant for a fraction of a minute after reaching a certain value, then continued to rise, while the helium depressurization continued to proceed linearly with time. For the duration of this plateau of the thermal output curve the slope of the elongation detector signal was significantly higher than before, but fell back to its original value after the plateau. This observation was made only for the reference rods. None of the annular rods, with or without prepressurization, nor the sphere-pac rods, showed the effect. When observed, the effect occurred at about 40 kw/m. The effect is attributed to fission gas release rapidly being enhanced by thermal feedback. The increase in stored energy associated with the temperature rise in the fuel causes the delay in thermal output. The larger available internal volume and/or the prepressurization of the annular rods, and the lack of a distinct fuel-clad gap for the sphere-pac rods prevented the effect from occurring in those other

Biological CO2 conversion to acetate in subsurface coal-sand formation using a high-pressure reactor system

Directory of Open Access Journals (Sweden)

Yoko eOhtomo

2013-12-01

Full Text Available Geological CO2 sequestration in unmineable subsurface oil/gas fields and coal formations has been proposed as a means of reducing anthropogenic greenhouse gasses in the atmosphere. However, the feasibility of injecting CO2 into subsurface depends upon a variety of geological and economic conditions, and the ecological consequences are largely unpredictable. In this study, we developed a new flow-through-type reactor system to examine potential geophysical, geochemical and microbiological impacts associated with CO2 injection by simulating in situ pressure (0–100 MPa and temperature (0–70°C conditions. Using the reactor system, anaerobic artificial fluid and CO2 (flow rate: 0.002 and 0.00001 mL/min, respectively were continuously supplemented into a column comprised of bituminous coal and sand under a pore pressure of 40 MPa (confined pressure: 41 MPa at 40°C for 56 days. 16S rRNA gene analysis of the bacterial components showed distinct spatial separation of the predominant taxa in the coal and sand over the course of the experiment. Cultivation experiments using sub-sampled fluids revealed that some microbes survived, or were metabolically active, under CO2-rich conditions. However, no methanogens were activated during the experiment, even though hydrogenotrophic and methylotrophic methanogens were obtained from conventional batch-type cultivation at 20°C. During the reactor experiment, the acetate and methanol concentration in the fluids increased while the δ13Cacetate, H2 and CO2 concentrations decreased, indicating the occurrence of homo-acetogenesis. 16S rRNA genes of homo-acetogenic spore-forming bacteria related to the genus Sporomusa were consistently detected from the sandstone after the reactor experiment. Our results suggest that the injection of CO2 into a natural coal-sand formation preferentially stimulates homo-acetogenesis rather than methanogenesis, and that this process is accompanied by biogenic CO2 conversion to

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

International Nuclear Information System (INIS)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A.

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U 3 O 8 and U 3 Si 2 dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U 3 O 8 -Al and five containing U 3 Si 2 -Al), with densities of 3.2 gU/cm 3 and 4.8 gU/cm 3 respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

User Manual for XnWlup2.0, A Software to Visualize Nuclear Data for Thermal Reactors in WIMS-D Libraries

International Nuclear Information System (INIS)

Thiyagarajan, T.K.; Ganesan, S.; Jagannathan, V.; Karthikeyan, R.

2002-10-01

A project to prepare an exhaustive handbook of WIMS-D cross sections for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully implemented. A computer software, called XnWlup2.0, with graphical user interface for MS Windows has been developed at BARC. This report summarizes the salient features of this new software for the users of WIMS-D libraries. Several sample outputs produced by the software are presented to illustrate the powerful use of this software for routine use in reactor physics analyses. (author)

Simultaneous Coproduction of Hydrogen and Ethanol in Anaerobic Packed-Bed Reactors

Directory of Open Access Journals (Sweden)

Cristiane Marques dos Reis

2014-01-01

Full Text Available This study evaluated the use of an anaerobic packed-bed reactor for hydrogen production at different hydraulic retention times (HRT (1–8 h. Two reactors filled with expanded clay and fed with glucose (3136–3875 mg L−1 were operated at different total upflow velocities: 0.30 cm s−1 (R030 and 0.60 cm s−1 (R060. The effluent pH of the reactors was maintained between 4 and 5 by adding NaHCO3 and HCl solutions. It was observed a maximum hydrogen production rate of 0.92 L H2 h−1 L−1 in R030 at HRT of 1 h. Furthermore, the highest hydrogen yield of 2.39 mol H2 mol−1 glucose was obtained in R060. No clear trend was observed by doubling the upflow velocities at this experiment. High ethanol production was also observed, indicating that the ethanol-pathway prevailed throughout the experiment.

Methyl mercury exposure in Swedish women with high fish consumption

International Nuclear Information System (INIS)

Bjoernberg, Karolin Ask; Vahter, Marie; Grawe, Kierstin Petersson; Berglund, Marika

2005-01-01

We studied the exposure to methyl mercury (MeHg) in 127 Swedish women of childbearing age with high consumption of various types of fish, using total mercury (T-Hg) in hair and MeHg in blood as biomarkers. Fish consumption was assessed using a food frequency questionnaire (FFQ), including detailed information about consumption of different fish species, reflecting average intake during 1 year. We also determined inorganic mercury (I-Hg) in blood, and selenium (Se) in serum. The average total fish consumption, as reported in the food frequency questionnaire, was approximately 4 times/week (range 1.6-19 times/week). Fish species potentially high in MeHg, included in the Swedish dietary advisories, was consumed by 79% of the women. About 10% consumed such species more than once a week, i.e., more than what is recommended. Other fish species potentially high in MeHg, not included in the Swedish dietary advisories, was consumed by 54% of the women. Eleven percent never consumed fish species potentially high in MeHg. T-Hg in hair (median 0.70 mg/kg; range 0.08-6.6 mg/kg) was associated with MeHg in blood (median 1.7 μg/L; range 0.30-14 μg/L; r s =0.78; p s =0.32; p s =0.37; p s =0.35; p=0.002, respectively). I-Hg in blood (median 0.24 μg/L; range 0.01-1.6 μg/L) increased with increasing number of dental amalgam fillings. We found no statistical significant associations between the various mercury species measured and the Se concentration in serum. Hair mercury levels exceeded the levels corresponding to the EPA reference dose (RfD) of 0.1 μg MeHg/kg b.w. per day in 20% of the women. Thus, there seems to be no margin of safety for neurodevelopmental effects in fetus, for women with high fish consumption unless they decrease their intake of certain fish species

Radiation doses and ground contamination in Sweden after a major nuclear reactor accident. An enquiry performed by the Swedish Radiation Protection Institute in concert with the Swedish Nuclear Power Inspectorate, September 1995; Straaldoser och markbelaeggning i Sverige efter en stor kaernkraftolycka. En utredning utfoerd av SSI i samraad med SKI, september 1995

Energy Technology Data Exchange (ETDEWEB)

Baeverstam, U

1995-12-01

Consequences of radioactive emissions from a hypothetical severe accident at a Swedish nuclear power plant are estimated. Three different cases are studied; two cases where the systems for reduction of accident consequences work properly; and one case where they don`t - a `worst` case. The first case, where the security systems are supposed to work fully, give limited consequences: between a few and about 50 cancer deaths in Europe (integrated in time) depending on wind directions. Food production would be affected in an area within 10 km from the reactor, but not to a large extent. The second case, where the security system do not function to 100%, but 0.1% of the total activity is released, would give 20-100 extra cancer deaths in the normally prevailing winds for all Swedish sites. Under very unfavourable wind conditions this sum may rise to 200, for the Barsebaeck site perhaps to 500. Ground contamination of Iodine can be heavy within short distance, with repercussions for agriculture. For the last, worst case, severe consequences may follow, possibly with acute radiation deaths in an area closer than 5 km from the reactor. In favourable wind conditions cancer deaths can amount to a few hundred over 50 years, at normal conditions up to 2000-8000 and in the most unfavourable weather perhaps twice that amount. Emergency evacuation would be recommended, under the plume, at a distance up to 100-150 km. This will however not be possible, due to lack of time. The high level of contamination will cause a long-time evacuation from the area within distances of up to 50 km. 45 refs, 8 tabs, 5 figs.

N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

Czech Academy of Sciences Publication Activity Database

Obalová, L.; Jirátová, Květa; Karásková, K.; Chromčáková, Ž.

2012-01-01

Roč. 191, č. 1 (2012), s. 116-120 ISSN 0920-5861 R&D Projects: GA TA ČR TA01020336 Institutional support: RVO:67985858 Keywords : N2O * catalytic decomposition * fixed bed reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.980, year: 2012

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

Summary of Swedish activities in the framework of the IWGATWR

International Nuclear Information System (INIS)

Pedersen, T.

1991-01-01

This summary starts with a brief review of the situation and outlook for nuclear power in Sweden from the political and industrial points of view, and to some extent from the public acceptance point of view. Then the Swedish activities in the field of advanced technologies for water-cooled reactors are outlined, the activities fall into three basic categories: activities related to operating plants, i.e. implementation of modern technology into these plants; development work on evolutionary type nuclear plants; and development work on more revolutionary or developmental type of reactors. Activities in the frameworks of the BWR 90 and PIUS projects are described. 3 figs, 1 tab

SKI - ASAR - R1. As operated Safety Analysis Report. Recurring safety review 1995 Ringhals 1

International Nuclear Information System (INIS)

2000-01-01

According to Swedish law, the reactor owner is responsible for performing a safety review and writing a so called ASAR-report. The Nuclear Power Inspectorate (SKI) examines this report, and reports the findings to the government (the so called SKI-ASAR-report). Each Swedish reactor should pass through three full ASAR reviews during its lifetime, similar to the licensing inspection before start-up of the reactor. The second series ASAR was delivered by the Ringhals utility to SKI in September 1995, and forms the basis for the SKI analysis in the present report

A pilot study of myofascial release therapy compared to Swedish massage in Fibromyalgia

Science.gov (United States)

Liptan, Ginevra; Mist, Scott; Wright, Cheryl; Arzt, Anna; Jones, Kim Dupree

2017-01-01

Summary Fibromyalgia (FM) is characterized by widespread muscle pain and soft tissue tenderness. However, a lack of definitive muscle pathology has made FM both a diagnostic and a treatment puzzle. Much of the evidence for pathology in FM lies in the central nervous system – in particular abnormal amplification of pain signals in the spinal cord – a manifestation of central sensitization. An emerging body of evidence posits that peripheral pain generated from the muscles and fascia may trigger and maintain central sensitization in FM. Since FM patients so frequently seek manual therapy to relieve muscle symptoms, the present study compared two different manual therapy techniques in a parallel study of women with FM. Eight subjects received myofascial release (MFR) while four subjects received Swedish massage, 90 min weekly for four weeks. Overall symptom burden and physical function were assessed by the Fibromyalgia Impact Questionnaire Revised (FIQ-R). A unique challenge for the manual therapist in treating conditions involving central sensitization is to determine if localized pain reduction can be achieved with targeted therapy in the context of ongoing widespread pain. Localized pain improvement was measured by a novel questionnaire developed for this study, the modified Nordic Musculoskeletal Questionnaire (NMQ). Between-group differences in FIQ-R did not reach statistical significance, but the total change scores on FIQ-R for the MFR group (mean = 10.14, SD = 16.2) trended in the hypothesized and positive direction compared to the Swedish massage group (mean = 0.33, SD = 4.93) yielding a positive Aikin separation test. Although overall modified NMQ scores improved in both groups there were no consistent focal areas of improvement for the Swedish massage group. In contrast, the MFR group reported consistent pain reductions in the neck and upper back regions on the NMQ. These data support the need for larger randomized controlled trials of MFR versus other

Decommissioning of Swedish nuclear power reactors. Technology and costs

International Nuclear Information System (INIS)

1994-06-01

The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

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)

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)

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Obtaining of total and thermal neutron flux in the carousel facility of the TRIGA MARK IPR-R1 reactor using the Monte Carlo transport method

International Nuclear Information System (INIS)

Guerra, Bruno Teixeira

2011-01-01

The IPR-R1 is a reactor type TRIGA, Mark-I model, manufactured by the General Atomic Company and installed at Nuclear Technology Development Centre (CDTN) of Brazilian Nuclear Energy Commission (CNEN), in Belo Horizonte, Brazil. It is a light water moderated and cooled, graphite-reflected, open-pool type research reactor. IPR-R1 works at 100 kW but it will be briefly licensed to operate at 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The fuel is an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The goal this work is modelling of the IPR-R1 Research Reactor TRIGA using the codes MCNPX2.6.0 (Monte Carlo N-Particle Transport extend) and MCNP5 to the calculating the neutron flux in the carousel facility. In each simulation the sample was placed in a different position, totaling forty positions around of the reactor core. The comparison between the results obtained with experimental values from other work showing a relatively good agreement. Moreover, this methodology is a theoretical tool in validating of the experimental values and necessary for determining neutron flux which can not be accessible experimentally. (author)

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

ORNL R and D on advanced small and medium power reactors: selected topics

International Nuclear Information System (INIS)

White, J.D.; Trauger, D.B.

1989-01-01

From 1984-1985, ORNL studied several innovative small and medium power nuclear concepts with respect to viability. Criteria for assessment of market attractiveness were developed and are described here. Using these criteria and descriptions of selected advanced reactor concepts, an assessment of their projected market viability in the time period 2000-2010 was made. All of these selected concepts could be considered as having the potential for meeting the criteria but, in most cases, considerable R and D would be required to reduce uncertainties. This work and later studies of safety and licensing of advanced, passively safe reactor concepts by ORNL are described. The results of these studies are taken into account in most of the current (FY 1989) work at ORNL on advanced reactors. A brief outline of this current work is given. One of the current R and D efforts at ORNL which addresses the operability and safety of advanced reactors is the Advanced Controls Program. Selected topics from this Program are described

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)

Biological biogas upgrading capacity of a hydrogenotrophic community in a trickle-bed reactor

International Nuclear Information System (INIS)

Rachbauer, Lydia; Voitl, Gregor; Bochmann, Günther; Fuchs, Werner

2016-01-01

Highlights: • Data on long term operation of a system supplied with real biogas are presented. • Ex-situ biological methanation is feasible for biogas upgrading. • Gas quality obtained complies with strictest direct grid injection criteria. • Biomethane can act as flexible storage for renewable surplus electricity. - Abstract: The current study reports on biological biogas upgrading by means of hydrogen addition to obtain biomethane. A mesophilic (37 °C) 0.058 m"3 trickle-bed reactor with an immobilized hydrogenotrophic enrichment culture was operated for a period of 8 months using a substrate mix of molecular hydrogen (H_2) and biogas (36–42% CO_2). Complete CO_2 conversion (> 96%) was achieved up to a H_2 loading rate of 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d, corresponding to 2.3 h gas retention time. The optimum H_2/CO_2 ratio was determined to be between 3.67 and 4.15. CH_4 concentrations above 96% were achieved with less than 0.1% residual H_2. This gas quality complies even with tightest standards for grid injection without the need for additional CO_2 removal. If less rigid standards must be fulfilled H_2 loading rates can be almost doubled (10.95 versus 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d) making the process even more attractive. At this H_2 loading the achieved methane productivity was 2.52 m_n"3 CH_4/m"3_r_e_a_c_t_o_r _v_o_l_. × d. In terms of biogas this corresponds to an upgrading capacity of 6.9 m_n"3 biogas/m"3_r_e_a_c_t_o_r _v_o_l_. × d. The conducted experiments demonstrate that biological methanation in an external reactor is well feasible for biogas upgrading under the prerequisite that an adequate H_2 source is available.

Study on magnetic properties of (Nd{sub 0.8}Ce{sub 0.2}){sub 2−x}Fe{sub 12}Co{sub 2}B (x = 0–0.6) alloys

Energy Technology Data Exchange (ETDEWEB)

Tan, G.S.; Xu, H., E-mail: huixu8888@shu.edu.cn; Yu, L.Y.; Tan, X.H.; Zhang, Q.; Gu, Y.; Hou, X.L.

2017-09-01

Highlights: • (Nd{sub 0.8}Ce{sub 0.2}){sub 2−x}Fe{sub 12}Co{sub 2}B alloys are prepared by melt-spinning method with simultaneously decreasing of Nd, Ce concentration. • The magnetic properties B{sub r}, (BH){sub max} and squareness are all improved with an appropriate reduction of Nd, Ce concentration. • Magnetic field heat treatment offers a significant improvement in B{sub r}, (BH){sub max} and squareness. - Abstract: In the present work, (Nd{sub 0.8}Ce{sub 0.2}){sub 2−x}Fe{sub 12}Co{sub 2}B (x = 0–0.6) permanent alloys are prepared by melt-spinning method. The hard magnetic properties of (Nd{sub 0.8}Ce{sub 0.2}){sub 2−x}Fe{sub 12}Co{sub 2}B (x = 0–0.6) alloys annealed at optimum temperatures have been investigated systematically. Depending on the Nd, Ce concentration, the maximum energy product ((BH){sub max}) and remanence (B{sub r}) increase gradually with x in the range of 0 ≤ x ≤ 0.4, whereas decrease gradually in the alloys with 0.4 < x ≤ 0.6. It is found that the optimum magnetic properties are obtained at x = 0.4: H{sub ci} = 4.9 kOe, B{sub r} = 10.1 kG, (BH){sub max} = 13.7 MGOe. Specifically, magnetic field heat treatment below the Curie temperature is applied for (Nd{sub 0.8}Ce{sub 0.2}){sub 1.6}Fe{sub 12}Co{sub 2}B (x = 0.4) annealed ribbons. The magnetic properties B{sub r}, (BH){sub max} and squareness are all enhanced after the magnetic field heat treatment. The (BH){sub max} shows a substantial increase from 13.7 MGOe to 16.0 MGOe after the heat treatment at 623 K with a magnetic field of 1 T, which gets 17% improvement compared with that of the sample without a magnetic field heat treatment. We demonstrate that the magnetic field heat treatment plays a certain role in the magnetization reversal behavior and can improve the microstructure of (Nd{sub 0.8}Ce{sub 0.2}){sub 1.6}Fe{sub 12}Co{sub 2}B alloy.

Strengthening the R and D on fast reactor technology, and promoting its industrialization

International Nuclear Information System (INIS)

Wan Gang

2008-01-01

Based on the strategic thoughts of energy development in China expounded by Jiang Zemin in the article entitled 'Reflections on Energy Issues in China', the author points out in this paper that R and Ds on fast reactor technology shall be carried out timely in China by taking full advantage of international scientific resources, and overall planning in this regard shall be made as well. The point of view of strengthening fast reactor technology R and D and promoting its industrialization is also put forward in the paper. (authors)

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

Application of nondestructive methods for qualification of high density fuels in the IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Jose E.R.; Silva, Antonio T.; Domingos, Douglas B.; Terremoto, Luis A.A., E-mail: jersilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN/CNEN-SP manufactures fuels to be used in its research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil still does not have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds (U{sub 3}O{sub 8} and U{sub 3}Si{sub 2} dispersed in Al matrix) internationally tested and qualified to be used in research reactors, and has attained experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans at IPEN/CNEN-SP to increase the uranium density of the fuels. Ten fuel miniplates (five containing U{sub 3}O{sub 8}-Al and five containing U{sub 3}Si{sub 2}-Al), with densities of 3.2 gU/cm{sup 3} and 4.8 gU/cm{sup 3} respectively, are being irradiated inside an irradiation device placed in a peripheral position of the IEA-R1 core. Non-destructive methods will be used to evaluate irradiation performance of the fuel miniplates after successive cycles of irradiation, by means: monitoring the reactor parameters during operation; periodic underwater visual inspection of fuel miniplates, eventual sipping test for fuel miniplates suspected of leakage and underwater measuring of the miniplate thickness for assessment of the fuel miniplate swelling. (author)

Translation and testing of measurement properties of the Swedish version of the IKDC subjective knee form.

Science.gov (United States)

Tigerstrand Grevnerts, H; Grävare Silbernagel, K; Sonesson, S; Ardern, C; Österberg, A; Gauffin, H; Kvist, J

2017-05-01

To translate to Swedish language and cross-culturally adapt the IKDC-SKF and to test the measurement properties of the Swedish version of IKDC-SKF in ACL-injured patients undergoing reconstruction surgery.The translation and cross-cultural adaption was performed according to guidelines. Seventy-six patients with an ACL injury filled out the IKDC-SKF and other questionnaires before ACL reconstruction and at 4, 6, and 12 months after surgery. A total of 203 patients from the Swedish ACL Registry participated at 8 months post-operative. Measurement properties were tested according to the COnsensus-based Standards for the selection of health Measurement INstruments (COSMIN) guidelines.The Swedish IKDC-SKF had high internal consistency (Cronbach's alpha=0.90) and test-retest reliability (ICC 2,1 =0.92, CI 95%: 0.81-0.97, Pmeasurement properties and can be recommended for use in a population of ACL-deficient patients undergoing ACL reconstruction. © 2017 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

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)

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)

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.

Safety approach and R and D program for future french sodium-cooled fast reactors

International Nuclear Information System (INIS)

Beils, Stephane; Carluec, Bernard; Devictor, Nicolas; Fiorini, Gian Luigi; Sauvage, Jean Francois

2011-01-01

This paper presents briefly the safety approach as well as the R and D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R and D. B) Strategy and roadmap in support of the R and D for future SFRs. This section describes the R and D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype. (author)

Alcohol use disorder and divorce: evidence for a genetic correlation in a population-based Swedish sample.

Science.gov (United States)

Salvatore, Jessica E; Larsson Lönn, Sara; Sundquist, Jan; Lichtenstein, Paul; Sundquist, Kristina; Kendler, Kenneth S

2017-04-01

We tested the association between alcohol use disorder (AUD) and divorce; estimated the genetic and environmental influences on divorce; estimated how much genetic and environmental influences accounted for covariance between AUD and divorce; and estimated latent genetic and environmental correlations between AUD and divorce. We tested sex differences in these effects. We identified twin and sibling pairs with AUD and divorce information in Swedish national registers. We described the association between AUD and divorce using tetrachorics and used twin and sibling models to estimate genetic and environmental influences on divorce, on the covariance between AUD and divorce and the latent genetic and environmental correlations between AUD and divorce. Sweden. A total of 670 836 individuals (53% male) born 1940-1965. Life-time measures of AUD and divorce. AUD and divorce were related strongly (males: r tet  = +0.44, 95% CI = 0.43, 0.45; females r tet  = +0.37, 95% CI = 0.36, 0.38). Genetic factors accounted for a modest proportion of the variance in divorce (males: 21.3%, 95% CI = 7.6, 28.5; females: 31.0%, 95% CI = 18.8, 37.1). Genetic factors accounted for most of the covariance between AUD and divorce (males: 52.0%, 95% CI = 48.8, 67.9; females: 53.74%, 95% CI = 17.6, 54.5), followed by non-shared environmental factors (males: 45.0%, 95% CI = 37.5, 54.9; females: 41.6%, 95% CI = 40.3, 60.2). Shared environmental factors accounted for a negligible proportion of the covariance (males: 3.0%, 95% CI = -3.0, 13.5; females: 4.75%, 95% CI = 0.0, 6.6). The AUD-divorce genetic correlations were high (males: rA = +0.76, 95% CI = 0.53, 0.90; females +0.52, 95% CI = 0.24, 0.67). The non-shared environmental correlations were modest (males: rE = +0.32, 95% CI = 0.31, 0.40; females: +0.27, 95% CI = 0.27, 0.36). Divorce and alcohol use disorder are correlated strongly in the Swedish population, and the heritability of divorce is consistent

Application of non-destructive methods for qualification of the U3O8-Al and U3Si2-Al dispersion fuels in the IEA-R1 Reactor

International Nuclear Information System (INIS)

Silva, Jose Eduardo Rosa da

2011-01-01

IPEN/CNEN-SP manufactures fuels to be used in its nuclear research reactor - the IEA-R1. To qualify those fuels, it is necessary to check if they have a good performance under irradiation. As Brazil doesn't have nuclear research reactors with high neutron fluxes, or suitable hot cells for carrying out post-irradiation examination of nuclear fuels, IPEN/CNEN-SP has conducted a fuel qualification program based on the use of uranium compounds, internationally tested and qualified to be used in research reactors, and has gotten experience in the technological development stages for the manufacturing of fuel plates, irradiation and non-destructive post-irradiation testing. Fuel elements containing low volume fractions of fuel in the dispersion were manufactured and irradiated successfully directly in the core of the IEA-R1. However, there are plans to increase the uranium density of these fuels. The objective of this thesis work was to study and to propose a set of non-destructive methods to qualify the dispersions fuels U 3 O 8 -Al e U 3 Si 2 -Al with high uranium density produced at IPEN/CNEN-SP. For that, the irradiation resources in the IEA-R1, and the application of non-destructive methods in the reactor pool available in the Institution were considered. The proposal is to specify, manufacture and irradiate fuel mini plates in IEA-R1 at the maximum densities, qualified internationally, and to monitor their general conditions during the period of irradiation, using non-destructive methods in the reactor pool. In addition to the non-destructive visual inspection and sipping methods, already used at the Institution, the infrastructure for dimensional sub-aquatic testing to evaluate the swelling of irradiated fuel mini plates was completed. The analyses of the results will provide means to assess and decide whether or not to continue with the irradiation of mini plates, until the desired burnup for the irradiation tests at IEA-R1 are reached. (author)

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)

Validation of a Swedish version of the short UPPS-P impulsive behavior scale among young adults

Directory of Open Access Journals (Sweden)

Benjamin Claréus

2017-12-01

Full Text Available The UPPS-P model of impulsivity proposes that impulsivity comprises five distinct facets—negative urgency, positive urgency, lack of premeditation, lack of perseverance, and sensation seeking. The UPPS-P Impulsive Behavior Scale has been used to measure these facets. The purpose of the current study was to develop and evaluate the psychometric properties of a Swedish version of the 20-item UPPS-P Impulsive Behavior Scale (SUPPS-P. The sample comprised 343 Swedish young adults (Mage=24.21, SD=2.01; 27% men, 2% other or undisclosed gender identity who answered a questionnaire including the SUPPS-P; Depression, Anxiety, and Stress Scale (DASS-21; and questions regarding their alcohol consumption and substance use. Confirmatory factor analysis supported a 5-factor, inter-correlated model, where each subscale of the SUPPS-P constitutes one latent variable. The convergent validity was established by replicating previously found correlations between the different impulsivity facets and depression, anxiety, frequency of alcohol consumption, and substance use. The internal consistency was acceptable for all the SUPPS-P subscales (Cronbach's α=0.65–0.78, McDonald's ω=0.65–0.79, except lack of perseverance (Cronbach's α=0.60, McDonald's ω=0.61. Thus, while the Swedish version of the SUPPS-P is suitable for assessing impulsivity in Swedish young adult samples, further research is needed to improve the psychometric properties of the lack of perseverance subscale. Keywords: Impulsivity, UPPS, UPPS-P, Swedish, Impulsive behavior scale

Swedish Disarmament Policy

OpenAIRE

2012-01-01

NPIHP Partners Host Conference on Swedish Disarmament Policy Dec 05, 2012 The Nuclear Proliferation International History Project is pleased to announce a conference on Swedish nuclear disarmament policy, organized and hosted by Stockholm University on 26 november 2012. Organized by Stockholm University Professor Thomas Jonter, Emma Rosengren, Goran Rydeberg, and Stellan Andersson under the aegis of the Swedish Disarmament Resaerch Project, the conference featured keynote addresses by Hans Bl...

Neutronic parameters characterization of the TRIGA IPR-R1 using scale 6.0 (KENO VI)

Energy Technology Data Exchange (ETDEWEB)

Faria, Victor; Miro, Rafael; Verdu, Gumersindo; Barrachina, Teresa [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM), Universitat Politecnica de València (Spain); Silva, Clarysson A. Mello da; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

KENO-VI is a Monte Carlo based transport code used to obtain the criticality of a nuclear system. A model built using this code in the SCALE6.0 software system was developed for the characterization of neutronic parameters of the IPR-R1 TRIGA research reactor. A comparison with experimental values and those calculated with a MCNP code model could be then attained with the purpose to validate this methodology. (author)

Neutronic parameters characterization of the TRIGA IPR-R1 using scale 6.0 (KENO VI)

International Nuclear Information System (INIS)

Faria, Victor; Miro, Rafael; Verdu, Gumersindo; Barrachina, Teresa; Silva, Clarysson A. Mello da; Pereira, Claubia

2011-01-01

KENO-VI is a Monte Carlo based transport code used to obtain the criticality of a nuclear system. A model built using this code in the SCALE6.0 software system was developed for the characterization of neutronic parameters of the IPR-R1 TRIGA research reactor. A comparison with experimental values and those calculated with a MCNP code model could be then attained with the purpose to validate this methodology. (author)

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)

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

Data Qualification Report For: Thermodynamic Data File, DATA0.YMP.R0 For Geochemical Code, EQ3/6?

International Nuclear Information System (INIS)

P.L. Cloke

2000-09-01

The objective of this work is to evaluate the adequacy of chemical thermodynamic data provided by Lawrence Livermore National Laboratory (LLNL) as DataO.ymp.ROA in response to an input request submitted under AP-3.14Q. This request specified that chemical thermodynamic data available in the file, Data0.com.R2, be updated, improved, and augmented for use in geochemical modeling used in Process Model Reports (PMRs) for Engineered Barrier Systems, Waste Form, Waste Package, Unsaturated Zone, and Near Field Environment, as well as for Performance Assessment. The data are qualified in the temperature range 0 to 100 C. Several Data Tracking Numbers (DTNs) associated with Analysis/Model Reports (AMR) addressing various aspects of the post-closure chemical behavior of the waste package and the Engineered Barrier System that rely on EQ316 outputs to which these data are used as input, are Principal Factor affecting. This qualification activity was accomplished in accordance with the AP-SIII.2Q using the Technical Assessment method. A development plan, TDP-EBS-MD-000044, was prepared in accordance with AP-2.13Q and approved by the Responsible Manager. In addition, a Process Control Evaluation was performed in accordance with AP-SV.1Q. The qualification method, selected in accordance with AP-SIII.2Q, was Technical Assessment. The rationale for this approach is that the data in File Data0.com.R2 are considered Handbook data and therefore do not themselves require qualification. Only changes to Data0.com.R2 required qualification. A new file has been produced which contains the database Data0.ymp.R0, which is recommended for qualification as a result of this action. Data0.ymp.R0 will supersede Data0.com.R2 for all Yucca Mountain Project (YMP) activities

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

Commissioning of the new heat exchanger for the research nuclear reactor IEA-R1

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo Jose Alvim de; Cassiano, Douglas Alves; Umbehaun, Pedro Ernesto; Carvalho, Marcos Rodrigues de; Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ajcastro@ipen.br; docass@gmail.com.br; umbehaun@ipen.br; carvalho@ipen.br; frajndli@ipen.br

2008-07-01

The Research Reactor IEA-R1 placed at IPEN/CNEN-SP is of the swimming pool type, light water moderated and with graphite reflectors, and was build and designed by Babcock and Wilcox Co. Start up operation was in September the 16{sup th}, 1957, being the first criticality for South Hemisphere. Although designed to operate at 5 MW, the IEA-R1 was operated until 2001 with 2 MW and was suitable for use in basic and applied research as well as the production of medical radioisotopes, industry and natural sciences applications. Due to a recent demand increase on radioisotopes in Brazil for medical diagnoses and therapies applications, IPEN /CNEN updated the IEA-R1 power to 5 MW and to work at continuous operation regime. Studies on the Ageing Management for the Research Reactor IEA-R1 were conducted according to IAEA procedures. As result of these studies critical components within the Ageing Management Program were identified. Also were made recommendations on the implementation of test scheduling and standardization procedures to organize data and documents. One of the main results was the need of monitoring the two heat exchangers, the two primary circuit pumps and the data acquisition system. During monitoring procedures, issues were observed on the IEA-R1 operation at 5 MW mainly due to the ageing of the Babcox and Wilcox TCA heat exchanger, and excessive vibrations at high flow rates on CBC's TCB heat exchanger. So, from 2005 on, it was decided to work with 3,5 MW and provide a new IESA heat exchanger with 5 MW capacity, to substitute the TCA heat exchanger. This work presents results on the commissioning of the new heat exchanger and compares against the values calculated in the IESA project. The results show that the IEA-R1 Reactor can be operated more safety and continuously at 5 MW with the new IESA heat exchanger. (author)

The radial distribution of the neutron field in the core of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Determination of the radial distribution of the thermal neutron field in the core of the Dalat reactor is done by the Cu foil activation method. The measured data are fitted by the least square method to determine several physical parameters of the reactor, as follows: 1. Buckling B{sub r}{sup 2}=(84.6{+-}5.5)10{sup -4}/cm{sup 2}. 2. The effective radius R{sub eff}=(27.6{+-}1.0)cm. 3. The extrapolation distance {lambda}=(8.7{+-}1.0)cm. 4. The unequal coefficient of the effective multiplication K{sub r}=1.77{+-}0.11. (author). 2 refs., 4 figs., 1 tab.

Technology and costs for decommissioning Swedish nuclear power plants; Teknik och kostnader foer rivning av svenska kaernkraftverk

Energy Technology Data Exchange (ETDEWEB)

Hedin, Gunnar; Gustavsson, Boerje [Westinghouse Electric Sweden AB, Vaesteraas (Sweden); Carlsson, Jan [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

2004-06-01

SKB has already performed three studies on available technology and approximate costs for decommissioning Swedish NPPs (years 1986, 1994 and 2000). The present report is an update of the year 2000 report with emphasis on areas that have been studied since the publication of that report. The report also gives a review of the technologies that have been chosen for decommissioning the Swedish reactors. The cost-estimation has also been updated and indexed to the present monetary situation. Areas in need for further studies are pointed in the report.

Radiation protection actions at Swedish nuclear power plants 1994-2002 and some reflections about the near future

International Nuclear Information System (INIS)

Erixon, Stig; Godaas, Tommy; Hofvander, Peter; Lund, Ingmar; Malmqvist, Lars; Thimgren, Ingela; Oelander-Guer, Hanna

2003-12-01

This report provides a summary of radiation protection experiences over the years 1994-2002 in the Swedish nuclear power industry. Actions to reduce radiation levels in reactor systems, occupational exposure results and some reflections about the near future are presented

Ultradense Nuclear Fusion in Metallic Lithium Liquid. A report on research performed at the R and D Center, Sakaguchi E.H VOC Co. under the auspices of the Swedish Energy Agency

International Nuclear Information System (INIS)

Ikegami, Hidetsugu; Pettersson, Roland

2006-10-01

This report is concerned with research and development on a new fusion scheme, 'chemonuclear fusion'. In this scheme, lithium or deuterium ions are implanted in liquid lithium whereby huge reaction rate enhancements, as much as up to 10 15 compared to what is expected for a free two-body deuterium-lithium interaction, are obtained. The enhancement is suggested to be a result of nuclear, atomic and chemical reactions taking place cooperatively. Experimental studies on the Li - D chemonuclear fusion is supported financially by the Swedish Energy Agency and were initiated at the Dept. of Analytical Chemistry, Uppsala University. The studies were continued in a collaboration with the R and D Centre, Sakaguchi E.H VOC Co. in Tokyo where a new and modified setup was constructed. Here, besides the Li - D chemonuclear fusion, the Li - Li fusion and the D 2 - 2Li molecular chemonuclear fusion were developed. In 2005 at the R and D Centre, molecular ions D 2 + of energies 30keV were implanted on a surface of metallic Li liquid. Product alpha particles were identified and measured by a single solid state detector. The energies were around 7.6MeV corresponding to what would be expected for the reaction 7 Li + D → 2x 4 He + n. Under some conditions of the Li liquid, the reaction rate was intermittently so high that the particle detector was saturated and stopped counting simultaneously with an appreciable temperature rise in the Li liquid. The results were discussed in March at the University of Tokyo and in October at the Royal Swedish Academy of Sciences, The Royal Swedish Engineering Academy of Sciences and at Uppsala University. This report presents a full description of the results. It also contains more recent results where an additional detector setup, a ΔE-E detector was used for validation of the results in particular the identity of the alpha particles

Study and application of k0-IAEA program on the Dalat Research Reactor

International Nuclear Information System (INIS)

Cao Dong Vu; Tran Quang Thien; Nguyen Thi Sy; Ho Manh Dung; Nguyen Nhi Dien

2011-01-01

The pneumatic 7-1 and rotary rack (lazy susan) facilities are two main irradiation channels for neutron activation analysis at Dalat research reactor. The experiments for characterizing these two irradiation facilities at June, July and August, 2010 by k 0 -IAEA using Al-0.1% Au, Al-0.1% Lu, 99.98% Ni and 99.8% Zr monitors were done. The neutron spectrum parameters include: (1) Thermal neutron flux (Φ th ); (2) The ratio between thermal neutron flux and epi-thermal neutron flux (f); (3) The α factor describing the neutron flux distribution 1/E 1+α , and (4) The neutron temperature (T n ). This study provides the first database for further studying and applying k 0 -IAEA at Dalat Research Reactor in the future. (author)

Evaluation of results from an in-pile creep test in the Studsvik R2 reactor

Energy Technology Data Exchange (ETDEWEB)

Pettersson, Kjell [Entropy Materials, Stockholm (Sweden)

2002-01-01

An in-pile creep test with bowing of cladding tubes has been performed in a hot water loop in the Studsvik R2 reactor . One test was performed in the core and one outside the core. The out-of-pile sample showed some minor primary creep strain while the in-pile specimen deformed at a steady rate of 5x10{sup -7}/h . However, when the results were compared to a broader data base of Zircaloy in-pile creep it became clear that the creep deformation observed is a primary creep which occurs before the irradiation creep in Zircaloy reaches a constant steady state creep rate. This primary stage is interpreted as a consequence of the development of an irradiation induced microstructure in Zircaloy which does not reach a steady state until a dose of about 10{sup 21} n/cm{sup 2} . At this stage the steady state irradiation creep starts. From this interpretation it is concluded that it is quite feasible to use the test method on pre-irradiated material in which it can be expected that the steady state will be reached already after short irradiation times.

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.

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)

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

Science.gov (United States)

Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

2017-12-01

The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

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)

The 5th surveillance testing for Kori unit 2 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2001-03-01

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 completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

Geosciences research: cooperation with Swedish Nuclear Fuel and Waste Management Co. (SKB)

International Nuclear Information System (INIS)

1993-01-01

PNC has been participating in the research program of the Construction Phase in Aespoe Hard Rock Laboratory project (HRL project), an underground research laboratory project initiated by Swedish Nuclear Fuel and Waste Management Company (SKB), since 1991. The main purpose of participating in the HRL project is to apply site characterization, prediction and validation methodology of geological environment in the project to R and D program on geological disposal in Japan. The outcome from investigations for the 0-700 m section in the access tunnel has been evaluated to compare with predictions on geological-structure. This report gives the summary of R and D program on the HRL project and preliminary results on evaluation of geological-structural predictions for the 0-700 m section in the access tunnel. (author)

Reliability Data Handbook for Piping Components in Nordic Nuclear Power Plants - R Book, Phase 2

International Nuclear Information System (INIS)

Hedtjaern Swaling, Vidar; Olsson, Anders

2011-02-01

This report presents results of a long research and development project financed by the regulatory body Straalsaekerhetsmyndigheten (SSM) (former SKI), the Swedish nuclear power plant licensees. The report presents a harmonized method for estimating Reliability Data for Piping Components in ASME code class 1 and 2 piping components (R-Book). Data in the R-Book is measured based on 'data driven' strategy. This first version of the R-Book comprises rupture frequencies and failure rates for all systems where ASME Code Class 1 or 2 events could be found in the OECD OPDE database. Nordic and Non-Nordic data are presented separately. Worldwide experience data is used to set up the relevant calculation cases, i.e. intersections of attributes for which there are at least one event present

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)

Pengaruh Swedish Massage Therapy terhadap Tingkat Kualitas Hidup Penderita Leukemia Usia Sekolah

Directory of Open Access Journals (Sweden)

Dewi Umu Kulsum

2017-08-01

Hasil penelitian menggambarkan terdapat perbedaan kualitas hidup pada kelompok intervensi sebelum dan sesudah dilakukan swedish massage therapy (p = 0,000 pada α = 5. Hasil penelitian merekomendasikan bahwa swedish massage therapy bisa dipakai sebagai metode alternatif dalam meningkatkan kualitas hidup penderita leukemia usia sekolah. Kata kunci: Kualitas hidup, leukemia, swedish massage therapy.

The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

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)

Recent plant studies using Victoria 2.0

International Nuclear Information System (INIS)

Bixler, Nathan E.; Gasser, Ronald D.

2000-01-01

VICTORIA 2.0 is a mechanistic computer code designed to analyze fission product behavior within the reactor coolant system (RCS) during a severe nuclear reactor accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS and secondary circuits. These predictions account for the chemical and aerosol processes that affect radionuclide behavior. VICTORIA 2.0 was released in early 1999; a new version VICTORIA 2.1, is now under development. The largest improvements in VICTORIA 2.1 are connected with the thermochemical database, which is being revised and expanded following the recommendations of a peer review. Three risk-significant severe accident sequences have recently been investigated using the VICTORIA 2.0 code. The focus here is on how various chemistry options affect the predictions. Additionally, the VICTORIA predictions are compared with ones made using the MELCOR code. The three sequences are a station blackout in a GE BWR and steam generator tube rupture (SGTR) and pump-seal LOCA sequences in a 3-loop Westinghouse PWR. These sequences cover a range of system pressures, from fully depressurized to full system pressure. The chief results of this study are the fission product fractions that are retained in the core, RCS, secondary, and containment and the fractions that are released into the environment

Steam explosions in light water reactors

International Nuclear Information System (INIS)

1981-01-01

The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

The Effect of Organic Loading Rate on Milk WastewaterTreatment Using Sequencing Batch Reactor (SBR

Directory of Open Access Journals (Sweden)

Hooman Hajiabadi

2009-09-01

Full Text Available In this study, four aerobic sequencing batch reactors (SBRs were operated under the same conditions for the treatment of milk wastewater at different organic loading rates (OLRs. Cylindrical Plexiglas reactors were run for 56 days (including 21 days of acclimatization and 35 days of data gathering. Effective volume, influent wastewater flowrate, and sludge retention time (SRT of reactors were 5.5 L, 3.5 L/d, and 10 d, respectively. The average COD removal efficiency for the reactors R1, R2, R3, and R4 with influent OLRave values of 633, 929, 1915, and 3261 gCOD/m3d were 95, 96, 95, and 82 percent, respectively. The average effluent suspended solid (SS for all reactors was lower than 44 mg/L. Also, except for R4 with an average effluent turbidity of 270 NTU, other reactors met the Iranian wastewater emission standard (50 NTU. In addition, the average sludge volume index of reactors R1 to R3 was found to be lower than 67 mL/g. According to the results, the overall variation of COD removal efficiency versus influent OLR shows a decreasing rate with a correlation factor of 0.8 (R2.

Prevalence of footrot in Swedish slaughter lambs

Directory of Open Access Journals (Sweden)

Nyman Ann-Kristin J

2011-04-01

Full Text Available Abstract Background Footrot is a world-wide contagious disease in sheep and goats. It is an infection of the epidermis of the interdigital skin, and the germinal layers of the horn tissue of the feet. The first case of footrot in Swedish sheep was diagnosed in 2004. Due to difficulties in distinguishing benign footrot from early cases of virulent footrot and because there is no possibility for virulence testing of strains of Dichelobacter nodosus in Sweden, the diagnosis is based of the presence or absence of clinical signs of footrot in sheep flocks. Ever since the first diagnosed case the Swedish Animal Health Service has worked intensively to stop the spread of infection and control the disease at flock level. However, to continue this work effectively it is important to have knowledge about the distribution of the disease both nationally and regionally. Therefore, the aims of this study were to estimate the prevalence of footrot in Swedish lambs at abattoirs and to assess the geographical distribution of the disease. Methods A prevalence study on footrot in Swedish lambs was performed by visual examination of 2000 feet from 500 lambs submitted from six slaughter houses. Each foot was scored according to a 0 to 5 scoring system, where feet with score ≥2 were defined as having footrot. Moreover, samples from feet with footrot were examined for Dichelobacter nodosus by culture and PCR. Results The prevalence of footrot at the individual sheep level was 5.8%, and Dichelobacter nodosus was found by culture and PCR in 83% and 97% of the samples from feet with footrot, respectively. Some minor differences in geographical distribution of footrot were found in this study. Conclusions In a national context, the findings indicate that footrot is fairly common in Swedish slaughter lambs, and should be regarded seriously.

Phasing out nuclear power, the swedish experience

International Nuclear Information System (INIS)

Fredriksson, Y.

2000-01-01

This article presents the chronological steps in the phasing-out of nuclear energy in Sweden. In 1980 a consultative referendum was held and it was decided that: i) no further expansion of nuclear capacity beyond the 12 reactors in operation or already under construction, ii) all nuclear power plants should be decommissioned by the year 2010. In 1988, as a consequence of the Chernobyl nuclear accident, the Swedish parliament decided that one reactor should be closed down in 1995 and a second in 1996. In 1991 the parliament proposed a new energy program for a 5 year period. The main measure was a huge financial support for increasing energy efficiency and for developing environmental sound technologies. At the same time the parliament repealed the 1991 decision of closing 1 reactor in 1995 and made the phase-out process dependent on the results of the new energy policy. In 1994 a parliamentary Commission was appointed to estimate the results of 1991 energy policy. The results were meager and disappointing so the Commission considered that a number of objectives (the climate issue, employment, welfare and competitiveness) remained unresolved if all nuclear power generation should be phased out by 2010. However, the Commission also considered it important to start the phasing-out process at an early stage and stated that one reactor could be closed down without noticeably affecting the power balance. The Barsebaeck reactor is to be closed before the end of november 1999. (A.C.)

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)

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)

Development of measuring system with self-powered neutron detectors for the LR-0 reactor

International Nuclear Information System (INIS)

Erben, O.; Horinek, K.; Szasz, Z.

1989-01-01

A measuring channel with self-powered detectors was developed for measuring neutron fluxs density in the reactor core. The measuring channel consists of a measuring probe with standard self-powered detectors of Soviet make, a signal pathway, a current/voltage converter and a measuring and recording unit. Neutron flux density in the LR-0 reactor core reaches a maximum of 10 13 m -2 s -1 . Experiments using the channel were carried out both in steady-state operation and after emergency shutdown of the reactor, this from power levels of 2,096 W and 1,830 W. The results of the experiments are tabulated and briefly analyzed. (Z.M.). 4 figs., 3 tabs., 5 refs

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)

The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

International Nuclear Information System (INIS)

Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

2010-01-01

LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

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)

Stage 2: dismantling of reactor case of the experimental F.B.R. Rapsodie

International Nuclear Information System (INIS)

Roger, J.

1994-01-01

This document defines the main objectives of stage 2 dismantling of the Rapsodie experimental fast neutron reactor and specifies its time schedule. The work already in progress consists in containing the reactor vessel and its internal equipment, as well as the neutron protection concrete, inside the two leak-tight barriers, and in dismantling all the systems and equipment systems contaminated by sodium. This work, which includes the destruction of 37 metric tons of contaminated sodium from the primary system, was begun in 1987 and will be completed in 1994. The duration of the waiting period for complete dismantling (stage 3) has not been defined. However, the containment and monitoring means implemented should allow a safe waiting period of several decades. (author). 4 figs

The evolution of doses in the IEA-R1 reactor environment and tendencies based on the current results

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio

2016-01-01

The IPEN / CNEN-SP have a Nuclear Research Reactor-NRR named IEA-R1, in operation from 1957. It is an open swimming pool reactor using light water as shielding, moderator and as cooling, the volume of this pool is 273m 3 .Until 1995 the reactor operated daily at a power of 2,0 MW. From June of that year, after a few safety modifications the reactor began operating in continuous way from Monday to Wednesday without shutdown totalizing 64 hours per week and the power was increased to 4,5MW also. Because of these changes, continuous operation and increased power, workers' doses would tend to increase. In the past several studies were conducted seeking ways to reduce the workers' doses. A study was made on the possibility to introduce a shielding at the top of the reactor core with a hot water layer. Studies have shown that a major limitation for operating a reactor at high power comes from the gamma radiation emitted by the sodium-24. Other elements such as magnesium-27, aluminum-28, Argon-51, contribute considerably to the water activity of the pool. The introduction of a hot water layer on the swimming pool would form a layer of surface, stable and free of radioactive elements with a 1.5m to 2m thickness creates a shielding to radiation from radioactive elements dissolved in water. Optimization studies proved that the installation of the hot layer was not necessary for the regime and the current power reactor operation, because other procedures adopted were more effective. From this decision the Radiological Protection Reactor Team, set up a dose assessment program to ensure them remained in low values based on principles established in national and international standards. The purpose of this paper is to analyze the individual doses of OEI (Occupationally Exposed Individual), which will be checked increasing doses resulting from recent changes in reactor operation regime and suggested viable safety and protection options, in the first instance to reducing

SSI's review of ASAR Ringhals 2, 1994

International Nuclear Information System (INIS)

Hofvander, P.

1995-11-01

Swedish nuclear power reactors are subject to periodic safety reviews, ASAR. Parts of ASAR deal with questions concerning radiation protection and are therefore submitted to a review performed by the Swedish Radiation Protection Institute. This report consists of SSI's review of ASAR Ringhals 2, 1994 . The following areas have been included in this review: Organisation, education, occupational exposures, effluents and discharges, emergency preparedness. 13 figs

Neutronic, thermal-hydraulics and accident analysis calculations for an irradiation device to be used in the qualification process of dispersion fuels in the IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas Borges; Silva, Antonio Teixeira e; Umbehaun, Pedro Ernesto; Silva, Jose Eduardo Rosa da; Conti, Thadeu das Neves; Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil)], e-mail: douglasborgesdomingos@yahoo.com.br

2009-07-01

Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of an irradiation device placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U{sub 3}O{sub 8}-Al e U{sub 3}Si{sub 2}-Al dispersion fuels, LEU type (19.9% of {sup 235}U), with uranium densities of, respectively, 3.0 gU/cm{sup 3} and 4.8gU/cm{sup 3}. The fuel miniplates will be irradiated to nominal {sup 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, now in the conception phase. For the neutronic calculation, the computer code CITATION was 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 codes LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. The calculations showed that the irradiation of the fuel miniplates will happen without any adverse consequence in the IEA-R1 reactor. (author)

Data Qualification Report For: Thermodynamic Data File, DATA0.YMP.R0 For Geochemical Code, EQ3/6 

Energy Technology Data Exchange (ETDEWEB)

P.L. Cloke

2001-10-16

The objective of this work is to evaluate the adequacy of chemical thermodynamic data provided by Lawrence Livermore National Laboratory (LLNL) as DataO.ymp.ROA in response to an input request submitted under AP-3.14Q. This request specified that chemical thermodynamic data available in the file, Data0.com.R2, be updated, improved, and augmented for use in geochemical modeling used in Process Model Reports (PMRs) for Engineered Barrier Systems, Waste Form, Waste Package, Unsaturated Zone, and Near Field Environment, as well as for Performance Assessment. The data are qualified in the temperature range 0 to 100 C. Several Data Tracking Numbers (DTNs) associated with Analysis/Model Reports (AMR) addressing various aspects of the post-closure chemical behavior of the waste package and the Engineered Barrier System that rely on EQ316 outputs to which these data are used as input, are Principal Factor affecting. This qualification activity was accomplished in accordance with the AP-SIII.2Q using the Technical Assessment method. A development plan, TDP-EBS-MD-000044, was prepared in accordance with AP-2.13Q and approved by the Responsible Manager. In addition, a Process Control Evaluation was performed in accordance with AP-SV.1Q. The qualification method, selected in accordance with AP-SIII.2Q, was Technical Assessment. The rationale for this approach is that the data in File Data0.com.R2 are considered Handbook data and therefore do not themselves require qualification. Only changes to Data0.com.R2 required qualification. A new file has been produced which contains the database Data0.ymp.R0, which is recommended for qualification as a result of this action. Data0.ymp.R0 will supersede Data0.com.R2 for all Yucca Mountain Project (YMP) activities.

Survey of nuclear parameters from the TRIGA Mark I IPR R1 Brazilian reactor with concentric configuration aiming the application of K0 neutron activation technique

International Nuclear Information System (INIS)

Franco, Milton Batista

2006-01-01

This research intended to determine the nuclear parameters a, f, spectral index and neutron temperature in several irradiations positions of the TRIGA Mark 1 IPR-R1 reactor, for use on the parametric method K 0 in the CDTN. K 0 is a monostandard method of neutron activation analysis. It is, on the whole, experimentally simple, flexible and an important tool for accurate and convenient standardization in instrumental multi-element analysis. At the time the parameters were determined at the rotatory rack, lower layer and in the central thimble: alpha was calculated applying the three bare monitor method using 197 Au, 94 Zr and 96 Zr; f determination was done according to the bare bi-isotopic method; neutron temperature was calculated through the direct method using 176 Lu, 94 Zr, 96 Zr and 197 Au and the Westcott's g(Tn) function for the 176 Lu was calculated and the result was interpolated in the Grintakis and Kim (1975) Table, determining the neutron temperature. The procedure to check the parameters consisted in using standard solutions of Au (metal foil, NBS), Lu (LuO 2 , Johnson Mattey Company - JMC) and Zr (ZrO 2 and metal foil, Johnson Mattey Company 99,99% and Zry - 4: 98,14% of Zr, National Bureau of Standard- NBS). Several certified reference materials and two samples of intercomparisons (samples of sediment of the IAEA/ARCAL XXVI project) have been analysed by means of k 0 - INAA in order to verify the efficiency of the method and the quality of the parameters. The certified reference materials were: GXR-2, GXR-5 and GXR-6 of the United States Geological Survey (USGS) and Soil-5, Soil-7 and SL-1 of the International Atomic Energy Agency (IAEA). (author)

Improved Dechlorinating Performance of Upflow Anaerobic Sludge Blanket Reactors by Incorporation of Dehalospirillum multivorans into Granular Sludge

Science.gov (United States)

Hörber, Christine; Christiansen, Nina; Arvin, Erik; Ahring, Birgitte K.

1998-01-01

Dechlorination of tetrachloroethene, also known as perchloroethylene (PCE), was investigated in an upflow anaerobic sludge blanket (UASB) reactor after incorporation of the strictly anaerobic, reductively dechlorinating bacterium Dehalospirillum multivorans into granular sludge. This reactor was compared to the reference 1 (R1) reactor, where the granules were autoclaved to remove all dechlorinating abilities before inoculation, and to the reference 2 (R2) reactor, containing only living granular sludge. All three reactors were fed mineral medium containing 3 to 57 μM PCE, 2 mM formate, and 0.5 mM acetate and were operated under sterile conditions. In the test reactor, an average of 93% (mole/mole) of the effluent chloroethenes was dichloroethene (DCE), compared to 99% (mole/mole) in the R1 reactor. The R2 reactor, with no inoculation, produced only trichloroethene (TCE), averaging 43% (mole/mole) of the effluent chloroethenes. No dechlorination of PCE was observed in an abiotic control consisting of sterile granules without inoculum. During continuous operation with stepwise-reduced hydraulic retention times (HRTs), both the test reactor and the R1 reactor showed conversion of PCE to DCE, even at HRTs much lower than the reciprocal maximum specific growth rate of D. multivorans, indicating that this bacterium was immobilized in the living and autoclaved granular sludge. In contrast, the R2 reactor, with no inoculation of D. multivorans, only converted PCE to TCE under the same conditions. Immobilization could be confirmed by using fluorescein-labeled antibody probes raised against D. multivorans. In granules obtained from the R1 reactor, D. multivorans grew mainly in microcolonies located in the centers of the granules, while in the test reactor, the bacterium mainly covered the surfaces of granules. PMID:9572963

Synthesis of nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} by absorption of ammonia into water-in-oil microemulsion in a rotor–stator reactor

Energy Technology Data Exchange (ETDEWEB)

Li, Yingwen; Wang, Hongrun; Arowo, Moses; Sun, Baochang, E-mail: sunbc@mail.buct.edu.cn; Chen, Jianfeng; Shao, Lei, E-mail: shaol@mail.buct.edu.cn [Beijing University of Chemical Technology, State Key Laboratory of Organic–Inorganic Composites (China)

2015-01-15

A gas-microemulsion reaction precipitation method was employed to prepare nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} by absorption of NH{sub 3} into water-in-oil (W/O) microemulsion in a rotor–stator reactor . The effects of different operating conditions including final pH of the microemulsion, reaction temperature, initial Ce{sup 3+} and Zr{sup 4+} concentration, rotation speed, and gas–liquid volumetric ratio were investigated. Nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} with an average diameter of about 5.5 nm, a specific surface area of 215.6 m{sup 2}/g and a size distribution of 4–8 nm was obtained under the optimum operating conditions. The as-prepared nano-Ce{sub 0.5}Zr{sub 0.5}O{sub 2} was loaded with Au to prepare nano-Au/Ce{sub 0.5}Zr{sub 0.5}O{sub 2} catalyst which was subsequently used for CO oxidation test. CO conversion rate reached 100 % at room temperature, indicating high catalytic activity of the nano-Au/Ce{sub 0.5}Zr{sub 0.5}O{sub 2} catalyst.

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.

Validation of a Swedish version of the short UPPS-P impulsive behavior scale among young adults.

Science.gov (United States)

Claréus, Benjamin; Daukantaitė, Daiva; Wångby-Lundh, Margit; Lundh, Lars-Gunnar

2017-12-01

The UPPS-P model of impulsivity proposes that impulsivity comprises five distinct facets-negative urgency, positive urgency, lack of premeditation, lack of perseverance, and sensation seeking. The UPPS-P Impulsive Behavior Scale has been used to measure these facets. The purpose of the current study was to develop and evaluate the psychometric properties of a Swedish version of the 20-item UPPS-P Impulsive Behavior Scale (SUPPS-P). The sample comprised 343 Swedish young adults ( M age  = 24.21, SD  = 2.01; 27% men, 2% other or undisclosed gender identity) who answered a questionnaire including the SUPPS-P; Depression, Anxiety, and Stress Scale (DASS-21); and questions regarding their alcohol consumption and substance use. Confirmatory factor analysis supported a 5-factor, inter-correlated model, where each subscale of the SUPPS-P constitutes one latent variable. The convergent validity was established by replicating previously found correlations between the different impulsivity facets and depression, anxiety, frequency of alcohol consumption, and substance use. The internal consistency was acceptable for all the SUPPS-P subscales (Cronbach's α = 0.65-0.78, McDonald's ω = 0.65-0.79), except lack of perseverance (Cronbach's α = 0.60, McDonald's ω = 0.61). Thus, while the Swedish version of the SUPPS-P is suitable for assessing impulsivity in Swedish young adult samples, further research is needed to improve the psychometric properties of the lack of perseverance subscale.

Big problems for Swedish nuclear industry

International Nuclear Information System (INIS)

Holmstroem, Anton; Runesson, Linda

2006-01-01

A report of the problems for Swedish nuclear industry the summer of 2006. A detailed description of the 25th of July incident at Forsmark 1 is provided. The incident was classified as level two on the INIS scale. The other Swedish nuclear plants were subject to security evaluations in the aftermath, and at Forsmark 2 similar weaknesses were found in the security system (ml)

RIA Analysis of Unprotected TRIGA Reactor

Directory of Open Access Journals (Sweden)

M.H. Altaf

2017-07-01

Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

Justify of implementation of a hot water layer system in swimming pool research reactor IEA-R1m

International Nuclear Information System (INIS)

Toyoda, Eduardo Yoshio; Gordon, Ana Maria Pinho Leite; Sordi, Gian-Maria A.A.

2001-01-01

The IPEN/CNEN-SP has a swimming pool research reactor (IEA-R1m) in operation since 1957 at 2 MW. In 1998, after some modifications, its nominal power increased to 5 MW. Among these modifications some adaptations had to be accomplished in the radiological protection and operational procedure. The present work aim to study the need of implementation of a hot water layer in order to reduce the dose in the workers in the vicinity of the reactor swimming pool. Applying the principles of radioprotection optimization, it was concluded that the decision of the construction of one hot water layer system in the reactor swimming pool, is not necessary. (author)

An approach to estimate the reactivity worth of R-5 poison tube system and experimental verification in ZERLINA reactor

International Nuclear Information System (INIS)

Khosla, S.K.; Paul, O.P.K.; Sengupta, S.N.

1976-01-01

It is proposed to employ a liquid poison injection system as an emergency shut down device for R-5 reactor. The liquid poison consists of gadolinium nitrate solution, which is injected into twenty poison tubes made of zircaloy that are located in between the regular lattice positions in R-5 reactor. The calculational model adopted to estimate the reactivity worth of the poison tubes so as to hold the reactor subcritical by 50 mk at full tank, is described. Similar reactivity estimates have also been carried out for R-5 poison tubes installed in Zerlina reactor in order to assess the adequacy of the calculational mode. The results of the calculations are compared with experimental values for single poison tubes. (author)

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

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

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

Energy Technology Data Exchange (ETDEWEB)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H., E-mail: songm@dgu.edu [Department of Mechanical, Robotics, and Energy Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of)

2015-03-15

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

International Nuclear Information System (INIS)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

2015-01-01

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made

ROMUSE 2.0 User Manual

Energy Technology Data Exchange (ETDEWEB)

Khuwaileh, Bassam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Turinsky, Paul [North Carolina State Univ., Raleigh, NC (United States); Williams, Brian J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-10-04

ROMUSE (Reduced Order Modeling Based Uncertainty/Sensitivity Estimator) is an effort within the Consortium for Advanced Simulation of Light water reactors (CASL) to provide an analysis tool to be used in conjunction with reactor core simulators, especially the Virtual Environment for Reactor Applications (VERA). ROMUSE is written in C++ and is currently capable of performing various types of parameters perturbations, uncertainty quantification, surrogate models construction and subspace analysis. Version 2.0 has the capability to interface with DAKOTA which gives ROMUSE access to the various algorithms implemented within DAKOTA. ROMUSE is mainly designed to interface with VERA and the Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design (SCALE) [1,2,3], however, ROMUSE can interface with any general model (e.g. python and matlab) with Input/Output (I/O) format that follows the Hierarchical Data Format 5 (HDF5). In this brief user manual, the use of ROMUSE will be overviewed and example problems will be presented and briefly discussed. The algorithms provided here range from algorithms inspired by those discussed in Ref.[4] to nuclear-specific algorithms discussed in Ref. [3].

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)

The low-energy spectrum of (2, 0) theory on T5 x R

International Nuclear Information System (INIS)

Henningson, Maans

2008-01-01

We consider the ADE-series of (2, 0) supersymmetric quantum theories on T 5 x R, where the first factor is a flat spatial five-torus, and the second factor denotes time. The quantum states of such a theory Φ are characterized by a discrete quantum number f element of H 3 (T 5 , C), where the finite abelian group C is the center subgroup of the corresponding simply connected simply laced Lie group G. At energies that are low compared to the inverse size of the T 5 , the spectrum consists of a set of continua of states, each of which is characterized by the value of f and some number 5r of additional continuous parameters. By exploiting the interpretation of this theory as the ultraviolet completion of maximally supersymmetric Yang-Mills theory on T 4 x S 1 x R with gauge group G adj = G/C and coupling constant g given by the square root of the radius of the S 1 factor, one may compute the number N f r (Φ) of such continua. We perform these calculations in detail for the A- and D-series. While the Yang-Mills theory formalism is manifestly invariant under the SL 4 (Z) mapping class group of T 4 , the results are actually found to be invariant under the SL 5 (Z) mapping class group of T 5 , which provides a strong consistency check.

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-01

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 ⁷LiF + BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in ⁷Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

Two Stage Anaerobic Reactor Design and Treatment To Produce Biogas From Mixed Liquor of Vegetable Waste

Science.gov (United States)

Budiastuti, H.; Ghozali, M.; Wicaksono, H. K.; Hadiansyah, R.

2018-01-01

Municipal solid waste has become a common challenged problem to be solved for developing countries including Indonesia. Municipal solid waste generating is always bigger than its treatment to reduce affect of environmental pollution. This research tries to contribute to provide an alternative solution to treat municipal solid waste to produce biogas. Vegetable waste was obtained from Gedebage Market, Bandung and starter as a source of anaerobic microorganisms was cow dung obtained from a cow farm in Lembang. A two stage anaerobic reactor was designed and built to treat the vegetable waste in a batch run. The capacity of each reactor is 20 liters but its active volume in each reactor is 15 liters. Reactor 1 (R1) was fed up with mixture of filtered blended vegetable waste and water at ratio of 1:1 whereas Reactor 2 (R2) was filled with filtered mixed liquor of cow dung and water at ratio of 1:1. Both mixtures were left overnight before use. Into R1 it was added EM-4 at concentration of 10%. pH in R1 was maintained at 5 - 6.5 whereas pH in R1 was maintained at 6.5 - 7.5. Temperature of reactors was not maintained to imitate the real environmental temperature. Parameters taken during experiment were pH, temperature, COD, MLVSS, and composition of biogas. The performance of reactor built was shown from COD efficiencies reduction obtained of about 60% both in R1 and R2, pH average in R1 of 4.5 ± 1 and R2 of 7 ± 0.6, average temperature in both reactors of 25 ± 2°C. About 1L gas produced was obtained during the last 6 days of experiment in which CH4 obtained was 8.951 ppm and CO2 of 1.087 ppm. The maximum increase of MLVSS in R1 reached 156% and R2 reached 89%.

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

14C emission from Swedish nuclear power plants and its effect on the 14C levels in the environment

International Nuclear Information System (INIS)

Stenstroem, K.; Erlandsson, Bengt; Hellborg, R.; Kiisk, M.; Persson, Per; Mattsson, Soeren; Thornberg, C.; Skog, G.

2000-02-01

The radionuclide 14 C is produced in all types of nuclear reactors mainly by neutron induced reactions in oxygen ( 17 O), nitrogen ( 14 N) and carbon ( 13 C). Part of the 14 C created is continuously released during normal operation as airborne effluents in various chemical forms (such as CO 2 , CO and hydrocarbons) to the surroundings. Because of the biological importance of carbon and the long physical half-life of 14 C, it is of interest to measure the releases and their incorporation into living material. The 14 C activity concentrations in annual tree rings and air around two Swedish nuclear power plants (Barsebaeck and Forsmark) as well as the background 14 C activity levels from two reference sites in southern Sweden during 1973-1996 are presented in this report. In order to verify the reliability of the method some investigations have been conducted at two foreign nuclear sites, Sellafield fuel reprocessing plant in England, and Pickering nuclear generating station in Canada, where the releases of 14 C are known to be substantial. Furthermore, results from some measurements in the vicinity of Paldiski submarine training centre in Estonia are presented. The results of the 14 C measurements of air, vegetation and annual tree rings around the two Swedish nuclear power plants show very low enhancements of 14 C, if at all above the uncertainty of the measurements. Even if the accuracy of the measurements of the annual tree rings is rather good (1-2%) the contribution of 14 C from the reactors to the environment is so small that it is difficult to separate it from the prevailing background levels of 14 C . This is the case for all sampling procedures: in air and vegetation as well as in annual tree rings. Only on a few occasions an actual increase is observed. However, although the calculations suffer from rather large uncertainties, the calculated release rate from Barsebaeck is in fair agreement with reported release data. The results of this investigation show

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

First Observations of {Upsilon}(1S) {r_arrow} {gamma}{pi}{sup +}{pi}{sup {minus}} and {Upsilon}(1S) {r_arrow} {gamma}{pi}{sup 0}{pi}{sup 0}

Energy Technology Data Exchange (ETDEWEB)

Anastassov, A.; Duboscq, J.E.; Gan, K.K.; Hart, T.; Honscheid, K.; Kagan, H.; Kass, R.; Lee, J.; Schwarthoff, H.; Wolf, A.; Zoeller, M.M. [Ohio State University, Columbus, Ohio 43210 (United States); Richichi, S.J.; Severini, H.; Skubic, P.; Undrus, A. [University of Oklahoma, Norman, Oklahoma 73019 (United States); Bishai, M.; Chen, S.; Fast, J.; Hinson, J.W.; Menon, N.; Miller, D.H.; Shibata, E.I.; Shipsey, I.P. [Purdue University, West Lafayette, Indiana 47907 (United States); Glenn, S.; Kwon, Y.; Lyon, A.L.; Roberts, S.; Thorndike, E.H. [University of Rochester, Rochester, New York 14627 (United States); Jessop, C.P.; Lingel, K.; Marsiske, H.; Perl, M.L.; Savinov, V.; Ugolini, D.; Zhou, X. [Stanford Linear Accelerator Center, Stanford University, Stanford, California 94309 (United States); Coan, T.E.; Fadeyev, V.; Korolkov, I.; Maravin, Y.; Narsky, I.; Stroynowski, R.; Ye, J.; Wlodek, T. [Southern Methodist University, Dallas, Texas 75275 (United States); Artuso, M.; Dambasuren, E.; Kopp, S.; Moneti, G.C.; Mountain, R.; Schuh, S.; Skwarnicki, T.; Stone, S.; Titov, A.; Viehhauser, G.; Wang, J.C. [Syracuse University, Syracuse, New York 13244 (United States); Bartelt, J.; Csorna, S.E.; McLean, K.W.; Marka, S.; Xu, Z. [Vanderbilt University, Nashville, Tennessee 37235 (United States); Godang, R.; Kinoshita, K.; Lai, I.C.; Pomianowski, P.; Schrenk, S. [Virginia Polytechnic Institute and State University, Blacksburg, Virginia 24061 (United States); Bonvicini, G.; Cinabro, D.; Greene, R.; Perera, L.P.; Zhou, G.J. [Wayne State University, Detroit, Michigan 48202 (United States); Chan, S.; Eigen, G.; Lipeles, E.; Miller, J.S.; Schmidtler, M.; Shapiro, A.; Sun, W.M.; Urheim, J.; Weinstein, A.J.; Wuerthwein, F. [California Institute of Technology, Pasadena, California 91125 (United States); Jaffe, D.E.; Masek, G.; Paar, H.P.; Potter, E.M.; Prell, S.; Sharma, V. [University of California, San Diego, La Jolla, California 92093 (United States); and others

1999-01-01

We report on a study of exclusive radiative decays of the {Upsilon}(1S) resonance collected with the CLEOthinspthinspII detector operating at the Cornell Electron Storage Ring. We present the first observation of the radiative decays {Upsilon}(1S){r_arrow}{gamma}{pi}{sup +}{pi}{sup {minus}} and {Upsilon}(1S){r_arrow}{gamma}{pi}{sup 0}{pi}{sup 0} . For the dipion mass regime m{sub {pi}{pi}}{gt}1.0 GeV , we obtain B({Upsilon}(1S){r_arrow}{gamma}{pi}{sup +}{pi}{sup {minus}})=(6.3{plus_minus}1.2{plus_minus} 1.3){times}10{sup {minus}5} and B({Upsilon}(1S){r_arrow}{gamma}{pi}{sup 0}{pi}{sup 0})=(1.7{plus_minus}0.6{plus_minus} 0.3){times}10{sup {minus}5} . {copyright} {ital 1999} {ital The American Physical Society }

Neutron activation analysis at CDTN/CNEN using the IPR-R1 Triga Mark I reactor

International Nuclear Information System (INIS)

Menezes, Maria Angela de B.C.; Maretti Junior, Fausto; Kastner, Geraldo Frederico; Amaral, Angela Maria; Souza, Wagner de

2009-01-01

This paper describes in summary the activities developed by the Laboratory for Neutron Activation Analysis since the starting up of the IPR-R1 TRIGA Mark I research reactor in 1960. This Laboratory is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for significant percentage of CDTN's analytical demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays the neutron activation analysis is responsible for 70% of the analytical demand and the k 0 - Instrumental method for 80% of this demand answering clients' request and researches. In Brazil, CDTN is the only Institute that fully masters the Instrumental Neutron Activation Analysis k 0 -method using its own nuclear reactor. (author)

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

Theory versus data: how to calculate R0?

Directory of Open Access Journals (Sweden)

Romulus Breban

2007-03-01

Full Text Available To predict the potential severity of outbreaks of infectious diseases such as SARS, HIV, TB and smallpox, a summary parameter, the basic reproduction number R(0, is generally calculated from a population-level model. R(0 specifies the average number of secondary infections caused by one infected individual during his/her entire infectious period at the start of an outbreak. R(0 is used to assess the severity of the outbreak, as well as the strength of the medical and/or behavioral interventions necessary for control. Conventionally, it is assumed that if R(0>1 the outbreak generates an epidemic, and if R(0<1 the outbreak becomes extinct. Here, we use computational and analytical methods to calculate the average number of secondary infections and to show that it does not necessarily represent an epidemic threshold parameter (as it has been generally assumed. Previously we have constructed a new type of individual-level model (ILM and linked it with a population-level model. Our ILM generates the same temporal incidence and prevalence patterns as the population-level model; we use our ILM to directly calculate the average number of secondary infections (i.e., R(0. Surprisingly, we find that this value of R(0 calculated from the ILM is very different from the epidemic threshold calculated from the population-level model. This occurs because many different individual-level processes can generate the same incidence and prevalence patterns. We show that obtaining R(0 from empirical contact tracing data collected by epidemiologists and using this R(0 as a threshold parameter for a population-level model could produce extremely misleading estimates of the infectiousness of the pathogen, the severity of an outbreak, and the strength of the medical and/or behavioral interventions necessary for control.

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Catalytic combustion of the retentate gas from a CO2/H2 separation membrane reactor for further CO2 enrichment and energy recovery

International Nuclear Information System (INIS)

Hwang, Kyung-Ran; Park, Jin-Woo; Lee, Sung-Wook; Hong, Sungkook; Lee, Chun-Boo; Oh, Duck-Kyu; Jin, Min-Ho; Lee, Dong-Wook; Park, Jong-Soo

2015-01-01

The CCR (catalytic combustion reaction) of the retentate gas, consisting of 90% CO 2 and 10% H 2 obtained from a CO 2 /H 2 separation membrane reactor, was investigated using a porous Ni metal catalyst in order to recover energy and further enrich CO 2 . A disc-shaped porous Ni metal catalyst, namely Al[0.1]/Ni, was prepared by a simple method and a compact MCR (micro-channel reactor) equipped with a catalyst plate was designed for the CCR. CO 2 and H 2 concentrations of 98.68% and 0.46%, respectively, were achieved at an operating temperature of 400 °C, GHSV (gas-hourly space velocity) of 50,000 h −1 and a H 2 /O 2 ratio (R/O) of 2 in the unit module. In the case of the MCR, a sheet of the Ni metal catalyst was easily installed along with the other metal plates and the concentration of CO 2 in the retentate gas increased up to 96.7%. The differences in temperatures measured before and after the CCR were 31 °C at the product outlet and 19 °C at the N 2 outlet in the MCR. The disc-shaped porous metal catalyst and MCR configuration used in this study exhibit potential advantages, such as high thermal transfer resulting in improved energy recovery rate, simple catalyst preparation, and easy installation of the catalyst in the MCR. - Highlights: • The catalytic combustion of a retentate gas obtained from the H 2 /CO 2 separation membrane. • A disc-shaped porous nickel metal catalyst and a micro-channel reactor for catalytic hydrogen combustion. • CO 2 enrichment up to 98.68% at 400 °C, 50,000 h −1 and H 2 /O 2 ratio of 2.

Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

International Nuclear Information System (INIS)

Dybeck, P.; Gustafsson, B.

1990-01-01

This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

Food intake and gestational weight gain in Swedish women.

Science.gov (United States)

Bärebring, Linnea; Brembeck, Petra; Löf, Marie; Brekke, Hilde K; Winkvist, Anna; Augustin, Hanna

2016-01-01

The objective of this study was to investigate if food intake (dairy, snacks, caloric beverages, bread, cheese, margarine/butter, potato/rice/pasta/grains, red meat, fish and fruit/berries/vegetables) is associated with gestational weight gain (GWG) in Swedish women. Four day food records from 95 pregnant Swedish women were collected in the last trimester. GWG was calculated as weighed body weight in the last trimester (median gestational week 36) minus self-reported pre-pregnancy body weight. Excessive GWG was defined according to the guidelines by the Institute of Medicine. Food groups tested for association with GWG were dairy (milk, yoghurt and sour milk), snacks (sweets, crisps, popcorn, ice cream and cookies, but not nuts and seeds), caloric beverages (soft drinks, juice, lemonade and non-alcoholic beer), bread, cheese, margarine/butter, potato/rice/pasta/grains, red meat, fish and fruit/berries/vegetables. Median (lower-upper quartiles) GWG was 12.1 kg (10.0-15.3). In total, 28 % had an excessive GWG. Excessive GWG was most common among pre-pregnancy overweight and obese women, where 69 % had an excessive GWG. Median daily intake of fruits and vegetables was 352 g (212-453), caloric beverages was 238 g (100-420) and snacks was 111 g (69-115). Multivariable linear regression analysis showed that intake of caloric beverages, snacks, fish, bread and dairy in the last trimester of pregnancy were positively related to GWG (R(2) = 0.32). Multivariable logistic regression analysis showed that intake of caloric beverages, snacks, fish, and bread was associated with higher odds ratios for excessive GWG. Intake of caloric beverages, snacks, fish and bread were positively related to excessive GWG. Thus, these results indicate that maternal dietary intake should be given higher attention in the antenatal care.

Status of the R and D activities on fast reactors and ADS in Brazil

International Nuclear Information System (INIS)

Maiorino, Jose Rubens

2001-01-01

Research and Development in Nuclear Science and Technology is conducted by Research Institutes of the Brazilian Nuclear Energy Commission. In Fast Reactor, R and D activities started in the sixties, and in 1972 a small Na loop (100 kW) was constructed. At the same time, during the seventies at IPEN, research in cooperation with GA for Gas Cooled Fast Breeder Reactor was conducted. The motivation of such research was Thorium Fuel Cycle. As a result of this research a Helium Loop was constructed and a Split Table Critical Assembly (ZPR) was designed. During the eighties, an agreement with ANSALDO-NIRA resulted in an acquisition of a Sodium Loop for Thermohydraulics studies, however it never had been assembled. At the same time, a concept of a Binary Breeder Reactor using two cycles, Th and U, was developed. During the nineties, a National Program to conduct R and D (pyroprocess; U-Zr Metallic Fuel; HT-9; Electromagnetic Pump; and a conceptual design of a Experimental Reactor (60/20 MWth/MWe)) was proposed, however it was closed at the end of the decade. Now, only academic research is being conducted, and it is summarized in this report. Basically, they are: an integral lead fast reactor concept for developing countries, and an alternative concept for a fast energy amplifier accelerator driven system. The first is an combination of best characteristics of the American Integral Fast Reactor and the Russian Lead Cooled Reactor. The second is a conceptual design of ADS helium cooled imbedded in a solid lead subcritical array of fuel, using more than one point of spallation trying to reduce the requirement for energy and current of the accelerator

Effect of effluent recycling and shock loading on the biodegradation of complex phenolic mixture in hybrid UASB reactors

Energy Technology Data Exchange (ETDEWEB)

Ramakrishnan, A.; Gupta, S.K. [Indian Institute of Technology, Bombay (India)

2008-06-15

This study describes the feasibility of anaerobic treatment of synthetic coal wastewater using four identical 13.5 L (effective volume) bench scale hybrid up flow anaerobic sludge blanket (HUASB) reactors (R1, R2, R3 and R4) under mesophilic (27 {+-} 5{sup o}C) conditions. Synthetic coal wastewater with an average chemical oxygen demand (COD) of 2240 mg/L and phenolics concentration of 752 mg/L was used as substrate. Effluent recirculation was employed at four different effluent to feed recirculation ratios (R/F) of 0.5, 1.0, 1.5 and 2.0 for 100 days to study the effect of recirculation on the performance of the reactors. Phenolics and COD removal was found to improve with increase in effluent recirculation. An effluent to feed recycle ratio of 1.0 resulted in maximum removal of phenolics and COD. Phenolics and COD removal improved from 88% and 92% to 95% each, respectively. The concentration of volatile fatty acids in the effluent was lower than the influent when effluent to feed recirculation was employed. Effect of shock loading on the reactors revealed that phenolics shock load up to 2.5 times increase in the normal input phenolics concentration in the form of continuous shock load for 4 days did not affect the reactors performance irreversibly.

Swedish Acceptance and Action Questionnaire (SAAQ): a psychometric evaluation.

Science.gov (United States)

Lundgren, Tobias; Parling, Thomas

2017-06-01

Psychological inflexibility and experiential avoidance are equivalent (with somewhat different connotations) concepts and refer to an unwillingness to remain in contact with particular private events. This concept is most often measured by the Acceptance and Action Questionnaire (AAQ-II) and is strongly related to psychopathology and behavioral effectiveness. In this study, the preliminary psychometric properties of the Swedish version of the AAQ-II (Swedish Acceptance and Action Questionnaire-SAAQ) are presented. The study is done in two steps. In the first step, the 10-item version of the AAQ-II is investigated through principal component analysis (n = 147). Secondly, due to problems with the component structure, the instrument is reduced to a six-item version and its validity and internal consistency are investigated (n = 154). The six-item version shows good concurrent and convergent validity as well as satisfying internal consistency (α = .85). Furthermore, the Swedish six-item version of the AAQ-II showed one strong component. Test-retest reliability was satisfactory (r = .80; n = 228). In future research, predictive and external validity would be important to investigate in order to further ensure that the SAAQ is a useful measure for clinical research. In conclusion, the SAAQ has satisfactory psychometric properties, but more data need to be gathered to further explore the possibilities for the instruments in Swedish contexts.

Real-time neutron radiography at the Iea-R1 m nuclear research reactor

International Nuclear Information System (INIS)

Menezes, M.O. de; Pugliesi, R.; Pereira, M.A.S.; Andrade, M.L.G.

2003-01-01

A LIXI (Light Intensifier X-ray Image) device has been employed in a real-time neutron radiography system. The LIXI is coupled to a video camera and the real-time images can be observed in a TV monitor, and processed in a computer. In order to get the real-time system operational, the neutron radiography facility installed at the IEA-R1 m nuclear research reactor of the IPEN-CNEN/S P has been optimized. The most important improvements were the neutron/gamma ratio, the effective energy of the neutron beam, decrease of the scattered radiation at the irradiation position, and the additional shielding of the video camera. Several one-frame as well as computer processed images are presented. The overall Modulation Transfer Function for the real-time system was obtained from the resolution parameter p = 0:44 +- 0:04 mm; the system sensitivity, evaluated for a Perspex step wedge, was determined and the average value is 0:70 +- 0:09 mm. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

International Nuclear Information System (INIS)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter

2015-01-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of 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. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

Design of a new wet storage rack for spent fuels from IEA-R1 reactor

Energy Technology Data Exchange (ETDEWEB)

Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of 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. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

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

Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

Modernization of the CDTN IPR-R1 TRIGA reactor instrumentation and control

International Nuclear Information System (INIS)

Mesquita, A.Z.; Costa, A.C.L.; Souza, R.M.G.P.

2009-01-01

The control system of the IPR-R1 was changed in 1995. Although since the year's 80 was generalized the use of microprocessor technology and video monitors for visual interface, in the IPR-R1 control room it was used analogical system by relay-based logic, and were maintained the mechanical strip chart recorders (ink-pen drive) to measure, monitor and store the operational parameters. It was maintained the measure and the control of, practically, the same variables of the original system, although the reactor power already have been upgraded to 100 kW and began the studies to increase it to 250 kW, which is the current core configuration. For 250 kW operations the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art instrumentation and control system using microprocessor technology is proposed to replace the present analogical systems. The new system can eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards as NRC and IEE specifications. This paper describes the research project in process in CDTN that has as objective the modernization of the IPR-R1 TRIGA reactor instrumentation and control of the operational variables. The project also will improve the accomplishment of neutronic and thermal-hydraulic experiments, foreseen in the CDTN research program. (author)

Cardiovascular Risk Factors in Patients with Addison's Disease: A Comparative Study of South African and Swedish Patients

Science.gov (United States)

Ross, Ian Louis; Bergthorsdottir, Ragnhildur; Levitt, Naomi; Dave, Joel Alex; Schatz, Desmond; Marais, David; Johannsson, Gudmundur

2014-01-01

Background Patients with Addison's disease (AD) in Scandinavia have an increased risk for premature death due to cardiovascular disease (CVD). Serum lipids are important risk factors for CVD and vascular mortality. Replacement doses of hydrocortisone have historically been higher in Sweden than South Africa. The primary aim was to study the lipid profiles in a large group of patients with AD with the hypothesis that the lipid profile in patients in Sweden would be worse than in South Africa. Methods In a cross-sectional study, 110 patients with AD (55 from South Africa, 55 from Sweden) matched for age, gender, ethnicity and BMI were studied. Anthropometric measures, blood pressure, lipids, highly sensitive C-reactive protein (hs-CRP) and adiponectin were studied. Results All patients were Caucasian and the majority were women N = 36 (65.5%). Mean (standard deviation; SD) ages of the Swedish and South African patients were 52.9 (13.0) and 52.6 (14.4) years and BMI 25.3 (3.2) and 25.8 (4.1) kg/m2, respectively. The mean total daily hydrocortisone dose was greater in the Swedish patients than the South African patients, [33.0 (8.1) versus 24.3 (8.0) mg; p<0.0001]. South African patients had higher median (interquartilerange; IQR) triglycerides (TG) [1.59 (1.1–2.46) versus 0.96 (0.74–1.6) mmol/l; p<0.001], total cholesterol (TC) [6.02(1.50) versus 5.13 (0.87) mmol/l; p<0.001], LDL-C [4.43 (1.44) versus 2.75 (0.80) mmol/l; p<0.001] and median hs-CRP [2.15 (0.93–5.45) versus 0.99 (0.57–2.10) mg/L; p<0.003] and lower HDL-C [0.80 (0.40) versus 1.86 (0.46) mmol/l; p<0.001] than the Swedish patients. Approximately 20% of the patients in both cohorts had hypertension and diabetes mellitus. Conclusions South African patients with AD have worse lipid profiles and higher hs-CRP compared to their matched Swedish patients, despite lower doses of hydrocortisone. It is uncertain at this time whether these are due to genetic or environmental factors. PMID:24603607

Cardiovascular risk factors in patients with Addison's disease: a comparative study of South African and Swedish patients.

Directory of Open Access Journals (Sweden)

Ian Louis Ross

Full Text Available BACKGROUND: Patients with Addison's disease (AD in Scandinavia have an increased risk for premature death due to cardiovascular disease (CVD. Serum lipids are important risk factors for CVD and vascular mortality. Replacement doses of hydrocortisone have historically been higher in Sweden than South Africa. The primary aim was to study the lipid profiles in a large group of patients with AD with the hypothesis that the lipid profile in patients in Sweden would be worse than in South Africa. METHODS: In a cross-sectional study, 110 patients with AD (55 from South Africa, 55 from Sweden matched for age, gender, ethnicity and BMI were studied. Anthropometric measures, blood pressure, lipids, highly sensitive C-reactive protein (hs-CRP and adiponectin were studied. RESULTS: All patients were Caucasian and the majority were women N = 36 (65.5%. Mean (standard deviation; SD ages of the Swedish and South African patients were 52.9 (13.0 and 52.6 (14.4 years and BMI 25.3 (3.2 and 25.8 (4.1 kg/m2, respectively. The mean total daily hydrocortisone dose was greater in the Swedish patients than the South African patients, [33.0 (8.1 versus 24.3 (8.0 mg; p<0.0001]. South African patients had higher median (interquartilerange; IQR triglycerides (TG [1.59 (1.1-2.46 versus 0.96 (0.74-1.6 mmol/l; p<0.001], total cholesterol (TC [6.02(1.50 versus 5.13 (0.87 mmol/l; p<0.001], LDL-C [4.43 (1.44 versus 2.75 (0.80 mmol/l; p<0.001] and median hs-CRP [2.15 (0.93-5.45 versus 0.99 (0.57-2.10 mg/L; p<0.003] and lower HDL-C [0.80 (0.40 versus 1.86 (0.46 mmol/l; p<0.001] than the Swedish patients. Approximately 20% of the patients in both cohorts had hypertension and diabetes mellitus. CONCLUSIONS: South African patients with AD have worse lipid profiles and higher hs-CRP compared to their matched Swedish patients, despite lower doses of hydrocortisone. It is uncertain at this time whether these are due to genetic or environmental factors.

Project Radiation Protection East. Swedish cooperation program for radiation protection in Eastern and Central Europe. Status Report, March 1996

International Nuclear Information System (INIS)

Snihs, J.O.; Johansson, Mai; Grapengiesser, S.; Bennerstedt, T.

1996-04-01

Until now the Swedish program for radiation protection work in central and Eastern europe has been granted 55 MSEK by the Swedish government. The projects are assessed, planned and performed in close cooperation with partner organizations in the East. Since 1994, radiation protection cooperation concerning the former Soviet Navy training reactors in Paldiski, Estonia, is included in Radiation Protection East. The government has granted 8 MSEK for this purpose. This report presents a summary over some 150 projects, their status, allocated funds and their distribution over countries and project areas. The presentation is updated up to March 1996. 7 figs

Delta 2.0

DEFF Research Database (Denmark)

Skott, Jeppe; Skott, Charlotte Krog; Jess, Kristine

DELTA 2.0 er en ny og helt opdateret udgave af Delta, der i ti år været brugt i matematiklærernes grund-, efter- og videreuddannelse. DELTA 2.0 er seriens almene fagdidaktik. Der er også fagdidaktiske overvejelser i de øvrige bøger i serien, men de er knyttet til specifikt matematisk indhold. DELTA...... 2.0 behandler mere generelle matematikdidaktiske problemstillinger såsom læringsteoretiske overvejelser i forbindelse med matematik, centrale aspekter af det at undervise i matematik og digitale teknologier som værktøj til at støtte elevers faglige læring af matematik....

Corporate Governance in the Swedish Banking Sector

OpenAIRE

Palmberg, Johanna

2010-01-01

This paper studies the corporate governance structure among Swedish banks. Who controls the Swedish banks and what characteristics does the Swedish banking sector have? Issues related to corporate governance such as ownership structure, board of directors and control-enhancing mechanisms will be studied. The Swedish banking law, how Swedish banks handled the financial crises and government measures to deal with the financial crisis is also analyzed.

Aspects of the Iea-R1 research reactor seismic evaluation

International Nuclear Information System (INIS)

Mattar Neto, Miguel

1996-01-01

Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

Swedish contribution to the greenhouse effect and required reductions to meet the 550 ppmv target

International Nuclear Information System (INIS)

Lindell, Lina; Nilsson, Kristina

2002-11-01

According to the Swedish Parliament, the Swedish international climate strategy should focus on a stabilisation of the concentration of greenhouse gases in the atmosphere. An equilibrium concentration lower than 550 ppmv CO 2 -equivalents should be achieved by the end of this century. As an interim target, the yearly emissions should not exceed 4.5 tonnes CO 2 -equivalents per capita by 2050. In this study an inventory of Swedish emissions from 1834 until 2000, for the six greenhouse gases regulated by the Kyoto Protocol, is carried out. Future emission scenarios for carbon dioxide during the time period 2000-2050 are also defined. This data is used for estimating the contribution to the greenhouse effect both today and in the future. Further it is investigated if the 2050-target is sufficient for not exceeding an atmospheric concentration of 550 ppmv. The required reduction for 2100 to reach an equilibrium concentration below this level is also estimated. The Swedish contribution to the greenhouse effect today is about 30 % larger than it should be according to the fairness factor used in this study. The Swedish emission target set for 2050 is sufficient for not exceeding 550 ppmv by that year. However, to reach a stabilisation of the concentration below this level the emissions have to be reduced to 1.0-1.5 tonnes CO 2 -equivalents per capita by 2100

Swedish contribution to the greenhouse effect and required reductions to meet the 550 ppmv target

Energy Technology Data Exchange (ETDEWEB)

Lindell, Lina; Nilsson, Kristina [Uppsala Univ. (Sweden). School of Engineering

2002-11-01

According to the Swedish Parliament, the Swedish international climate strategy should focus on a stabilisation of the concentration of greenhouse gases in the atmosphere. An equilibrium concentration lower than 550 ppmv CO{sub 2}-equivalents should be achieved by the end of this century. As an interim target, the yearly emissions should not exceed 4.5 tonnes CO{sub 2}-equivalents per capita by 2050. In this study an inventory of Swedish emissions from 1834 until 2000, for the six greenhouse gases regulated by the Kyoto Protocol, is carried out. Future emission scenarios for carbon dioxide during the time period 2000-2050 are also defined. This data is used for estimating the contribution to the greenhouse effect both today and in the future. Further it is investigated if the 2050-target is sufficient for not exceeding an atmospheric concentration of 550 ppmv. The required reduction for 2100 to reach an equilibrium concentration below this level is also estimated. The Swedish contribution to the greenhouse effect today is about 30 % larger than it should be according to the fairness factor used in this study. The Swedish emission target set for 2050 is sufficient for not exceeding 550 ppmv by that year. However, to reach a stabilisation of the concentration below this level the emissions have to be reduced to 1.0-1.5 tonnes CO{sub 2}-equivalents per capita by 2100.

The water desalination complex based on ABV-type reactor plant

International Nuclear Information System (INIS)

Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

1997-01-01

A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

Science.gov (United States)

2012-05-03

... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

2005-01-01

The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

Energy Technology Data Exchange (ETDEWEB)

Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

2011-02-15

The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

IEA-R1 reactor core simulation with RELAP5 code

International Nuclear Information System (INIS)

Rocha, Ricardo Takeshi Vieira da; Belchior Junior, Antonio; Andrade, Delvonei Alves de; Sabundjian, Gaiane; Umbehaum, Pedro Ernesto; Torres, Walmir Maximo

2005-01-01

This paper presents a preliminary RELAP5 model for the IEA-R1 core. The power distribution is supplied by the neutronic code, CITATION. The main objective is to model the IEA-R1 core and validate the model through the comparison of the results to the ones from COBRA and PARET, which were used in the Final Safety Analysis Report (FSAR) for this plant. Preliminary calculations regarding some simulations are presented. Boundary conditions are simulated through time dependent components. Results obtained are compared to those available for the IEA-R1. This study will be continued considering a model for the whole plant. Important transient and accidents will be analysed in order to verify the Emergency Core Cooling System - ECCS efficiency to hold its function as projected to preserve the integrity of the reactor core and guarantee its cooling. (author)

SKI - ASAR - B1/B2 As operated Safety Analysis Report. Recurring safety review 1995 Barsebaeck 1 and 2

International Nuclear Information System (INIS)

1996-12-01

According to Swedish law, the reactor owner is responsible for performing a safety review and writing a so called ASAR-report. The Nuclear Power Inspectorate (SKI) examines this report, and reports the findings to the government (the so called SKI-ASAR-report). Each Swedish reactor should pass through three full ASAR reviews during its lifetime, similar to the licensing inspection before start-up of the reactor. The second series ASAR was delivered by the Barsebaeck utility to SKI in September 1995, and forms the basis for the SKI analysis in the present report

Safety and radiation protection at the Swedish nuclear power plants 2000

International Nuclear Information System (INIS)

2001-04-01

During 2000 no events occurred, or discoveries were made, that seriously affected the reactor safety at the Swedish nuclear plants. The basic safety strategy is designed so that hidden faults and deficiencies shall not lead to any serious consequences for the plants. It is of outmost importance that the safety work at the plants is performed with the best effort and quality in order to realize this strategy. Especially in the new economic situation of the utilities after deregulation of the electricity market. The total radiation dose to the personnel and contracted workers at the plants was the lowest ever recorded with all NPPs running (8.1 man Sv). Corrosion damages led to a stand-still of two reactors during a long period, and thorough analyses were performed before the Inspectorate allowed a restart

Di-μ-iodido-bis{[(R-(+-2,2′-bis(diphenylphosphanyl-1,1′-binaphthyl-κ2P,P′]copper(I} 0.67-hydrate

Directory of Open Access Journals (Sweden)

Daniel Volz

2012-04-01

Full Text Available The structure of the title compound, [Cu2I2(C44H32P22]·0.67H2O, has been determined because of its interesting catalytic and optical features. The molecule, which has non-crystallographic C2-symmetry, consists of a core structure of two CuI ions, bridged by two iodide ions. Each CuI ion is also coordinated by one equivalent of the chiral bidentate (R-BINAP ligand [BINAP = 2,2′-bis(diphenylphosphanyl-1,1′-binaphthyl]. Thus, both cations show a distorted tetrahedral geometry being surrounded by two I atoms and two P atoms from the (R-BINAP ligands. The complex consists of isolated butterfly-shaped molecules featuring an angle of 146.11 (2° between adjacent CuI2 planes. The structure displays intramolecular C—H...I hydrogen bonding and contains disordered water. The absolute configuration of this chiral complex was determined by anomalous dispersion effects.