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

International Nuclear Information System (INIS)

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

1991-11-01

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

OECD Halden reactor project

International Nuclear Information System (INIS)

1974-01-01

A major part of the current research programme is devoted to irradiation experiments with a wide variety of heavily instrumented test fuel assemblies, in order to study the thermal and mechanical behavior of fuel rods through in-core measurements, in particular various forms of deformation of cladding and fuel as related to operational conditions and fuel rod design parameters. From these measurements mathematical models are being developed to explain quantitatively the deformation behavior, as well as the thermal properties of the fuel. During 1974, fifty-six instrumented fuel assemblies were irradiated in these experiments. Another major part of the Halden programme is aimed at the development and demonstration of advanced computer-based methods for plant and reactor core control, for safety and protection, and for overall supervision of nuclear power stations. Both the control methods themselves and the associated measurement and control apparatus are being elaborated, and during the year particular progress was made with the ''OPCOM'' process operator communication system

Modelization Post-test experiment IFA-650.10 HALDEN with FRAP series codes; Modelizacion post-test del experimento HALDEN IFA-650.10 con los codigos de la serie FRAP

Energy Technology Data Exchange (ETDEWEB)

Vallejo, I.; Herranz, L. E.

2013-07-01

There is a need to review the criteria for security relating to LOCA accidents , including the effect of different materials of pod, as well as conditions of high burned as fuel. In this work is modeled with code FRAPTRAN-1.4 the IFA-650.10 experiment executed in the experimental reactor HALDEN. It is an approximation to the thermo-hydraulic rod-refrigerant and the results are compared with experimental measurements. The thermal behavior shows good agreement with the experimental measures; mechanical parameters are observed light quality deviations in pod and very good quantitative agreement in the maximum elongation; the diameter calculated at the end of the simulation above - predicts the post-irradiation values and oxide presents a good deal.

OECD Halden Reactor Project

International Nuclear Information System (INIS)

1988-01-01

The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

Halden reactor project

International Nuclear Information System (INIS)

1980-01-01

The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

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

Directory of Open Access Journals (Sweden)

Takahiro Minato

2018-05-01

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

20. Annual report. OECD Halden reactor project. 1979

International Nuclear Information System (INIS)

1981-01-01

This is the Twentieth Annual Report on the OECD Halden Reactor Project, describing activities during 1979, the first year of the 1979-1981 Halden Agreement. Research work at the project is focussed on three areas: 1) In-core behaviour of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2) Prediction, surveillance and control of fuel and core performance, for which models of fuel and core behaviour are developed. 3) Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

Lessons learned in process control at the Halden Reactor Project

International Nuclear Information System (INIS)

Kennedy, W.G.

1989-12-01

This report provides a list of those findings particularly relevant to regulatory authorities that can be derived from the research and development activities in computerized process control conducted at the Halden Reactor Project. The report was prepared by a staff member of the US Nuclear Regulatory Commission working at Halden. It identifies those results that may be of use to regulatory organizations in three main areas: as support for new requirements, as part of regulatory evaluations of the acceptability of new methods and techniques, and in exploratory research and development of new approaches to improve operator performance. More than 200 findings arranged in nine major categories are presented. The findings were culled from Halden Reactor Project documents, which are listed in the report

In-core materials testing under LWR conditions in the Halden reactor

International Nuclear Information System (INIS)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

2002-01-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

In-core materials testing under LWR conditions in the Halden reactor

Energy Technology Data Exchange (ETDEWEB)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

2002-07-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

Human machine interaction research experience and perspectives as seen from the OECD Halden Reactor Project

International Nuclear Information System (INIS)

Oewre, F.

1999-01-01

In this paper a short review is given on important safety issues in the field of human machine interaction as expressed by important nuclear organisations such as USNRC, IAEA and the OECD NEA. Further on, a presentation is offered of research activities at the OECD Halden Reactor Project in the field of human machine interaction aiming to clarify some of the issues outlined by the above mentioned organisations. The OECD Halden Reactor Project is a joint undertaking of national nuclear organisations in 19 countries sponsoring a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency. One of the research areas is the man-machine systems research addressing the operator tasks in a control room environment. The overall objective is to provide a basis for improving today's control rooms through introduction of computer-based solutions for effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. (author)

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

Science.gov (United States)

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

2013-02-01

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

Integral approach to innovative fuel and material investigations in the Halden reactor

International Nuclear Information System (INIS)

Volkov, B.

2009-01-01

Integral approach used for fuel and material investigations in the Halden reactor can be used in support of qualification and certification of fuel to be introduced in commercial NPPs. This approach has been partly used for WWER fuel investigation in the Halden Reactor in a series of irradiation tests. In-pile fuel performance tests with reliable measurements provided by Halden instrumentation under different conditions can be used for validation of the WWER fuel behaviour models and verification of fuel performance codes. These models and codes can be used for qualification of innovative fuel behaviour under extended conditions

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

OECD Halden reactor project

International Nuclear Information System (INIS)

1978-01-01

This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

Benchmark Calculations on Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Ek, Mirkka; Kekkonen, Laura; Kelppe, Seppo; Stengaard, J.O.; Josek, Radomir; Wiesenack, Wolfgang; Aounallah, Yacine; Wallin, Hannu; Grandjean, Claude; Herb, Joachim; Lerchl, Georg; Trambauer, Klaus; Sonnenburg, Heinz-Guenther; Nakajima, Tetsuo; Spykman, Gerold; Struzik, Christine

2010-01-01

The assessment of the consequences of a loss-of-coolant accident (LOCA) is to a large extent based on calculations carried out with codes especially developed for addressing the phenomena occurring during the transient. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have not only generated a need to re-examine the LOCA safety criteria and to verify their continued validity, but also to confirm that codes show an appropriate performance especially with respect to high burnup phenomena influencing LOCA fuel behaviour. As part of international efforts, the OECD Halden Reactor Project program implemented a test series to address particular LOCA issues. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The Halden LOCA series, using high burnup fuel segments, contains test cases well suited for checking the ability of LOCA analysis codes to predict or reproduce the measurements and to provide clues as to where the codes need to be improved. The NEA Working Group on Fuel Safety, WGFS, therefore decided to conduct a code benchmark based on the Halden LOCA test series. Emphasis was on the codes' ability to predict or reproduce the thermal and mechanical response of fuel and cladding. Before starting the benchmark, participants were given the opportunity to tune their codes to the experimental system applied in the Halden LOCA tests. To this end, the data from the two commissioning runs were made available. The first of these runs went

Description of the quality system and the organisation of activities at the Halden Reactor Project

International Nuclear Information System (INIS)

Tiseth, Ann Katrine

1996-04-01

The Halden Reactor Project is a joint undertaking of national organisations in 19 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency. The quality assurance routines in force at the Halden Reactor Project are based on long term experience and are devised to ensure product quality and meeting of programme goals, customers expectations and authority requirements. Quality is an overall connotation related to all activities carried out at the Project and concerns all individuals performing such activities. It is also related to subcontractors and suppliers of parts and components. The results of the work depend very strongly on the quality level of each component, regardless how small it is. The activities at the Halden Reactor Project are organised in eight technical divisions, one advisory group and one administrative group. This report describes the quality system, the organisation of the activities in the divisions and the technical and administrative infrastructures. The quality system is built according to the international standards ISO 9001 and ISO 9000-3. Iso 9000-3 is ISO 9001 applied to software development, supply and maintenance. (author)

3. Halden Reactor Project Workshop

International Nuclear Information System (INIS)

Louka, Michael N.

2005-09-01

A workshop was held in Halden 2nd-3rd March 2005 to discuss 'VR in the Future Industrial Workplace: Working Together - Regardless of Distance'. The workshop sessions and discussions focused on design, operations and maintenance, training, and engineering virtual reality systems, and provided useful insights into the current state of the art of research and development in the fields of virtual and augmented reality. (Author)

Halden project activities on software dependability

International Nuclear Information System (INIS)

Dahll, G.; Sivertsen.

1994-01-01

Since 1977, the OECD Halden Reactor Project has been working in the field of software dependability. Special emphasis has been put on the use of software in safety critical systems. All phases in software development, from specification through software development, verification, and validation have been covered and are discussed in this article

I and C safety research at the OECD Halden reactor project

International Nuclear Information System (INIS)

Gran, B.A.

2007-01-01

The overall objective of the Halden Reactor Project research on software systems dependability is to contribute to the successful introduction of digital I and C systems into NPPs. When celebrating the 50 years of the Halden Project in 2008, about 100 written reports have been delivered within this research. This research covers a number of topics covering safety, reliability, validation and verification, quality assurance, risk assessment, requirement engineering, error propagation, qualitative and quantitative assessment. In the paper some activities are described, pinpointing the importance of good joint projects with organisations in the member countries

OECD Halden Reactor Project

International Nuclear Information System (INIS)

1983-01-01

The OECD Halden Reactor Project is both the oldest and the only one still in operation of the three major joint undertakings established at the inception of the OECD Nuclear Energy Agency. This publication has been printed in connection with its twenty-fifth anniversary as an international project. After presentation of the history and organization of the project, a thorough description of the past and present activities in the field of fuel performance and process control and surveillance is given. The projects's fuel testing programme is now focuessed on an investigation to define safety margins under normal operations as well as under various kinds of accident situations. Fuel research is also concerned with the characterisation of long term effects with regard to efficiency, operational safety and mapping of reliability and durability in the case of accidents with loss of coolant. In the field of process control and surveillance, research work is directly linked to the use of computers and colour graphics as tools in the control room. A fullscale simulator-based model and experimental control room has been constructed. The first experiments to be carried out in this laboratory will investigate the advantage of analysing alarms before they are presented to the operator. (RF)

A summary of lessons learned activities conducted at the OECD Halden Reactor Project

International Nuclear Information System (INIS)

Hallbert, B.P.

1997-01-01

A series of lessons learned studies have been conducted at the OECD Halden Reactor Project. The purpose of these lessons learned reports are to summarize knowledge and experience gained across a number of research project. This paper presents a summary of main issues addressed in four of these lessons learned projects. These are concerned with software development and quality assurance, software reliability, methods for test and evaluation of developed systems, and the evaluation of system design features

The post irradiation examination of three fuel rods from the IFA 429 experiment irradiated in the Halden Reactor

International Nuclear Information System (INIS)

Williams, J.

1979-11-01

A series of fuel rod irradiation experiments were performed in the Halden Heavy Boiling Water Reactor in Norway. These were designed to provide a range of fuel property data as a function of burn-up. One of these experiments was the IFA-429. This was designed to study the absorption of helium filling gas by the UO 2 fuel pellets, steady state and transient fission gas release and fuel thermal behaviour to high burn-up. This data was to be obtained as a function of fuel density, fuel grain size, initial fuel/cladding gap, average linear heat rating, burn-up and overpower transients. All the fuel is in the form of pressed and sintered UO 2 pellets enriched to 13 weight percent 235 U. All the rods were clad in Zircaloy 4 tube. The details of the experiment are given. The post irradiation examination included: visual examination, neutron radiography, dimensional measurements, gamma scanning, measurement of gases in fuel rods and internal free volume, burn-up analysis, metallographic examination, measurement of retained gas in UO 2 pellets, measurement of bulk density of UO 2 . The results are given and discussed. (U.K.)

Role of Halden Reactor Project for world-wide nuclear energy development

Energy Technology Data Exchange (ETDEWEB)

McGrath, M.A.; Volkov, B.

2011-07-01

The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

Role of Halden Reactor Project for world-wide nuclear energy development

International Nuclear Information System (INIS)

McGrath, M.A.; Volkov, B.

2011-01-01

The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

2008-10-15

A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

Influence of FGR complexity modelling on the practical results in gas pressure calculation of selected fuel elements from Dukovany NPP

International Nuclear Information System (INIS)

Lahodova, M.

2001-01-01

A modernization fuel system and advanced fuel for operation up to the high burnup are used in present time in Dukovany NPP. Reloading of the cores are evaluated using computer codes for thermomechanical behavior of the most loaded fuel rods. The paper presents results of parametric calculations performed by the NRI Rez integral code PIN, version 2000 (PIN2k) to assess influence of fission gas release modelling complexity on achieved results. The representative Dukovany NPP fuel rod irradiation history data are used and two cases of fuel parameter variables (soft and hard) are chosen for the comparison. Involved FGR models where the GASREL diffusion model developed in the NRI Rez plc and standard Weisman model that is recommended in the previous version of the PIN integral code. FGR calculation by PIN2k with GASREL model represents more realistic results than standard Weisman's model. Results for linear power, fuel centre temperature, FGR and gas pressure versus burnup are given for two fuel rods

The Halden Reactor Project workshop on HAMMLAB 2000

International Nuclear Information System (INIS)

Sebok, Angelia L.; Grini, Rolf-Einar; Larsen, Marit; Ness, Eyvind; Soerensen, Aimar

1998-01-01

A workshop on HAMMLAB 2000 was organised in Halden, May 26-27, 1997. The purpose of the workshop was to discuss and make recommendations on requirements for the design of HAMMLAB 2000 and to discuss the future research agenda. The workshop began with several presentations summarising the status of the current HAMMLAB 2000 project. Three invited speakers with expertise in human-machine laboratories and simulators delivered presentations of their experiences. Later, the workshop was divided into five working groups that discussed the following issues in parallel: (1) Technical Studies, Research Agenda; (2) Human Factors Research Agenda; (3) Data Collection; (4) Synergy with Other Industries; (5) Virtual Reality. Each group produced specific recommendations that were summarised by the group's facilitator in a joint session of the workshop. This report summarises the presentation of the invited speakers, and the discussions and recommendations of the individual working groups. (author)

The Halden Reactor Project workshop meeting on control room development

International Nuclear Information System (INIS)

Miberg, Ann Britt; Green, Marie; Haukenes, Hanne; Larsen, Marit; Seim, Lars Aage; Veland, Oeystein

1999-03-01

The 'Control Room Development' workshop was organised in. Halden, November 5-6, 1998. The purpose of the workshop was to bring forward recommendations for the future use of HAMMLAB with respect to control room development. The workshop comprised thirteen presentations summarising current issues and status in control room development projects and related projects. Following the presentations, five working groups were formed. The purpose of the working groups was to establish a set of recommendations for the future use of HAMMLAB. Each working group developed a set of recommendations. The outcomes of the working groups' discussions were summarised in plenum by the working group chairs. During the workshop, all participants excluding the Halden Project staff were asked to fill in a questionnaire indicating which research topics they found most interesting to pursue in future HAMMLAB research. The purpose of this report is to summarise the workshop participants' presentations, the working groups' discussions, and the recommendations given by the workshop participants concerning the future use of HAMMLAB (author) (ml)

IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

International Nuclear Information System (INIS)

Turnbull, J.A.

1996-01-01

Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

International Nuclear Information System (INIS)

Horvath, M. I.; Jenssen, H. K.; Munoz-Reja, C.; Tverberg, T.

2011-01-01

Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP), HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD-Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as partner in the fuel research programs. Various experiments are developed and performed also by providing materials, ENUSA collaborates with HRP on various experiments investigating the fuel behaviour, especially on Gd-bearing fuel. 20 years of successful collaboration between HRP and ENUSA is continuing with promising and results to ensure and enhance the safe operation of the Spanish and all other NPPs in the world. (Author) 12 refs.

An overview of the fuels and materials testing programme at the OECD Halden Reactor Project

Energy Technology Data Exchange (ETDEWEB)

Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

1997-08-01

The fuels and materials testing programme of the OECD Halden Reactor Project is aimed at investigations of fuel and cladding properties at high burnup, water chemistry effects and in-core materials ageing problems. For the execution of this programme, different types of irradiation rigs and experimental facilities providing typical power reactors conditions are available. Data are obtained from in-core sensors developed at the Halden Project; these are shortly described. An overview of the current test programme and the scope of the following years are briefly presented. (author). 5 refs, 3 figs.

Safety Significance of the Halden IFA-650 LOCA Test Results

International Nuclear Information System (INIS)

Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Petit, Marc; Hozer, Zoltan; Kelppe, Seppo; Khvostov, Grigori; Hafidi, Biya; Therache, Benjamin; Heins, Lothar; Valach, Mojmir; Voglewede, John; Wiesenack, Wolfgang

2010-01-01

The safety criteria for loss-of-coolant accidents were defined to ensure that the core would remain coolable. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have generated a need to re-examine these criteria and to verify their continued validity. As part of international efforts to this end, the OECD Halden Reactor Project program implemented a LOCA test series. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, IRSN, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The fourth test of the series, IFA-650.4 conducted in April 2006, caused particular attention in the international nuclear community. The fuel used in the experiment had a high burnup, 92 MWd/kgU, and a low pre-test hydrogen content of about 50 ppm. The test aimed at and achieved a peak cladding temperature of 850 deg. C. The rod burst occurred at 790 deg. C. The burst caused a marked temperature increase at the lower end and a decrease at the upper end of the system, indicating that fuel relocation had occurred. Subsequent gamma scanning showed that approximately 19 cm of the fuel stack were missing from the upper part of the rod and that fuel had fallen to the bottom of the capsule. PIE at the IFE-Kjeller hot cells corroborated this evidence of substantial fuel relocation. The fact that fuel dispersal could occur upon ballooning and burst, i.e. at cladding temperatures as low as 800 deg. C and thus far lower than the temperature entailed by the current 1200 deg. C / 17% ECR limit, caused concern. The

OECD: Halden reactor project

International Nuclear Information System (INIS)

1979-01-01

The work at the Project has continued in the two main fields: test fuel irradiation and fuel research, and computer based process supervision and control. Organizations participating in the Project continue to have their fuel irradiated in the Halden Reactor in instrumented test assemblies designed and manufactured by the Project. The Project's fuel studies continue to focus on specific subjects such as fuel pellet/cladding interaction and heat transfer, fission product release and fuel behavior under loss of coolant conditions. The work on process control and supervision continues in the highly relevant fields of core control and operator-process communication. A system for predictive core control is being developed while special mathematical methods for core power distribution control are being studied. Operator-process communication studies comprise use of computer simulation on colour display as important ingredients, while the work on developing a system for interactive plant disturbance analysis continues

Overview of the OECD Halden reactor project

International Nuclear Information System (INIS)

Vitanza, C.

2000-01-01

The OECD Halden Reactor Project is an international network dedicated to enhancing the safety and reliability of nuclear power plants. The project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the project's goals and of its technical infrastructure. The paper also contains a brief overview of results from the ongoing programme and of the main issues contemplated for the next three-year programme period (2000-2002). (author)

Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

Energy Technology Data Exchange (ETDEWEB)

Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Karlsen, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

Implementation of MOAS II diagnosis system at the OECD Halden Reactor project

International Nuclear Information System (INIS)

Kim, I.S.; Grini, R.E.; Nilsen, S.

1995-01-01

MOAS II is a surveillance and diagnosis system that uses several techniques for knowledge acquisition and diagnostic reasoning, e.g., goal tree-success tree, simplified directed graphs, diagnosis trees, and detailed knowledge of the process, such as mass or energy balance. This new approach was used at the Halden Man-Machine Laboratory of the OECD Halden Reactor Project. The performance of MOAS II, developed in G2 real-time expert system shell for the high-pressure preheaters of the NORS process, was tested against a variety of transient scenarios, including failures of control valves and sensors, and leakage of tubes of the preheaters. These tests showed that MOAS II successfully carried out its intended functions, i.e., quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The insights gained during the implementation are discussed

OECD Halden reactor project

International Nuclear Information System (INIS)

1977-01-01

The activities of the OECD Halden Reactor Project for the year 1975 are summarized. The period under review is the last year of the three year joint programme which commenced on 1st January, 1973. The main items reported upon are: process supervision and control, test fuel irradiation and fuel research, reactor operations, and administration and finance. The process supervision and control work has been concentrated in two fields: methods development for core surveillance and control, and systems development for operator-process communication. As for fuel test, investigations of the densification phenomenon have continued through irradiations to a maximum of about 16000MWd/tUO 2 . Axial and radial deformations of fuel rods are studied, with the effect of power transients upon the dimensional stability of fuel rods, and fuel-cladding heat transfer and fuel temperature. Thermal models for steady state and transient heat transfer in fuel rods have been developed and the work on thermomechanical models of claddings shows considerable promise

Overview of the OECD-Halden reactor project

International Nuclear Information System (INIS)

Vitanza, Carlo

2001-01-01

The OECD Halden Reactor Project is an international network dedicated to enhanced safety and reliability of nuclear power plants. The Project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the Project goals and of its technical infrastructure. The paper contains also a brief overview of results from the programme carried out in the time period 1997-1999 and of the main issues contemplated for the 3-year programme period 2000-2002

Evaluating usability of the Halden Reactor Large Screen Display. Is the Information Rich Design concept suitable for real-world installations?

International Nuclear Information System (INIS)

Braseth, Alf Ove

2013-01-01

Large Screen Displays (LSDs) are beginning to supplement desktop displays in modern control rooms, having the potential to display the big picture of complex processes. Information Rich Design (IRD) is a LSD concept used in many real-life installations in the petroleum domain, and more recently in nuclear research applications. The objectives of IRD are to provide the big picture, avoiding keyhole related problems while supporting fast visual perception of larger data sets. Two LSDs based on the IRD concept have been developed for large-scale nuclear simulators for research purposes; they have however suffered from unsatisfying user experience. The new Halden Reactor LSD, used to monitor a nuclear research reactor, was designed according to recent proposed Design Principles compiled in this paper to mitigate previously experienced problems. This paper evaluates the usability of the Halden Reactor LSD, comparing usability data with the replaced analogue panel, and data for an older IRD large screen display. The results suggest that the IRD concept is suitable for use in real-life applications from a user experience point of view, and that the recently proposed Design Principles have had a positive effect on usability. (author)

In-pile test results of HANA claddings in Halden research reactor

Energy Technology Data Exchange (ETDEWEB)

Baek, Jong Hyuk; Choi, Byoung Kwon; Jeong, Yong Hwan; Jung, Yun Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

It is a kind of facing tasks in the nuclear industry to develop advanced claddings for high burn-up fuel which is safer and more economical than the existing conventional ones. Since 1997, taking an initiative in KAERI, the Zr cladding development team has carried out the R and D activities for the development of the advanced claddings to be used in the high burn-up fuel (>70,000 MWD.MTU). The team had produced the advanced claddings (HANA, High-performance Alloy for Nuclear Application) from the patented composition and manufacturing process in the international collaboration with U.S. and Japan. Now, the HANA claddings have being demonstrated their good performances from the out-of-pile tests including the corrosion, creep, burst, tensile, microstructures LOCA, RIA, wear, and so on. In parallel to the out-of-pile performance tests, the HANA claddings are being undertaken to evaluate their in-pile properties in Halden research reactor. In this study, it is included the test overviews, conditions, and results of the HANA claddings in the Halden reactor.

Verification of HELIOS-MASTER system through benchmark of Halden boiling water reactor (HBWR)

International Nuclear Information System (INIS)

Kim, Ha Yong; Song, Jae Seung; Cho, Jin Young; Kim, Kang Seok; Lee, Chung Chan; Zee, Sung Quun

2004-01-01

To verify the HELIOS-MASTER computer code system for a nuclear design, we have been performed benchmark calculations for various reactor cores. The Halden reactor is a boiling, heavy water moderated reactor. At a full power of 18-20MWt, the moderator temperature is 240 .deg. C and the pressure is 33 bar. This study describes the verification of the HELIOS-MASTER computer code system for a nuclear design and the analysis of a hexagonal and D 2 O moderated core through a benchmark of the Halden reactor core. HELIOS, developed by Scandpower A/S, is a two-dimensional transport program for the generation of group cross-sections, and MASTER, developed by KAERI, is a three-dimensional nuclear design and analysis code based on the two-group diffusion theory. It solves the neutronics model with the TPEN (Triangle based Polynomial Expansion Nodal) method for a hexagonal geometry

The Halden Reactor Project Workshop on Studies of Operator Performance During Night Shift's

International Nuclear Information System (INIS)

Morisseau, Dolores S.; Braarud, Per Oeyvind; Collier, Steve; Droeivoldsmo, Asgeir; Larsen, Marit; Lirvall, Peter

1996-01-01

A workshop on Studies of Operator Performance during Nights Shifts was organised in Halden, February 27-28, 1996. The purpose of the workshop was to discuss and make recommendations on specific needs for the study of operator cognitive performance at night and identify the relevant research issues for which Halden could provide resolution. The workshop began with presentations by several invited speakers with expertise in studies of shift work and was then divided into three working groups that discussed the following issues in parallel: (1) Lines of Research to Be Pursued; (2) Methods and Measures to Be Used in Research on Cognitive Performance at Nights; and (3) Products of the Research on Operator Performance at Night. Each group produced specific recommendations that were summarised by the group's facilitator in a joint session of the workshop. This report summarises the presentation of the invited speakers, and the discussions and recommendations of the individual working groups. (author)

Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

International Nuclear Information System (INIS)

1999-01-01

Academy of Sciences, KFKI Atomic Energy Research Institute, the N.V. KEMA, the Netherlands, the Russian Research Centre 'Kurchatov Institute', the Slovakian VUJE - Nuclear Power Plant Research Institute, and from USA: the ABB Combustion Engineering Inc., the Electric Power Research Institute (EPRI), and the General Electric Co. The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given this right by one of these Project member organisations. The activities in the area of fuel and materials performance are based on extensive in-reactor measurements. The programmes are expanding in the areas of fuel performance at extended burn-ups, waterside corrosion and material testing in general. Development of in-core instruments is an important activity in support of the experimental programmes. The research programme at the Halden Project addresses the research needs of the nuclear industry in connection with introduction of digital I and C systems in NPPs. The programme provides information supporting design and licensing of upgraded, computer-based control room systems, and demonstrates the benefits of such systems through validation experiments in Halden's experimental research facility, HAMMLAB and pilot installations in NPPs. The Enlarged Halden Programme Group Meeting at Loen, Norway, was arranged to provide an opportunity to present results of work carried out at Halden and within participating organisations, and to encourage comments and impulses related to future Halden Project work. This HPR-351 relates to the fuel and materials part of the meeting and is divided in two volumes, HPR-351 Volume I and HPR-351 Volume II. The corresponding collection of papers in the man-machine area are given in one volume, HPR-352 Volume I. The overall programme of the Loen Enlarged Meeting covering the Fuel and Materials Research is given in the following pages. The papers with denomination HWR have

Performance of MOX fuel: An overview of the experimental programme of the OECD Halden Reactor Project and review of selected results

International Nuclear Information System (INIS)

Wiesenack, W.; McGrath, M.

2000-01-01

The OECD Halden Reactor Project has defined an extensive experimental programme related to MOX fuels which is being executed with the objective to provide a performance data base similar to that available for UO 2 . In addition to utilising fresh MOX fuel and re-instrumented segments from LWR irradiations to high burnup, the concept of inert matrix fuel is being addressed. The irradiation in the Halden reactor is performed in rigs allowing steady state, power ramping and cyclic operation. In-pile data are obtained from instrumentation such as fuel centreline thermocouples, pressure transducers, fuel and cladding elongation detectors, and movable gauges for measuring the diametral deformation. Various phenomena can be assessed in this way, e.g. thermal performance, swelling and densification, PCMI and fission gas release. The paper describes the objectives of various experiments and provides examples of temperature, pressure and cladding elongation measurements performed on MOX fuel. Salient results are related to the threshold for the onset of significant fission gas release and the relaxation behaviour in a power ramp-PCMI situation. (author)

The Halden Reactor Project workshop meeting on human centred automation and function allocation methods

International Nuclear Information System (INIS)

Sebok, Angelia; Green, Marit; Larsen, Marit; Miberg, Ann Britt; Morisseau, Dolores

1998-02-01

A workshop on Human Centred Automation (HCA) and Function Allocation Methods was organised in Halden, September 29-30, 1997. The purpose of the workshop was to discuss and make recommendations on requirements for the Halden Project research agenda. The workshop meeting began with several presentations summarising current issues in HCA, Function Allocation Methods and Functional Modelling. Invited speakers presented their research or modelling efforts. Following the presentations, the workshop was divided into three working groups, all tasked with answering the same four questions: (1) What are the most important issues in Human Centred Automation? (2) Which strengths could be achieved by integrating Functional Modelling Methods into experimental Human Centred Automation research? (3) How should analytical and experimental methods be balanced? (4) What are the most important aspects in automation design methodology? Each group discussed the questions and produced specific recommendations that were summarised by the group's facilitator in a joint session of the workshop. (author)

Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

International Nuclear Information System (INIS)

Wiesenack, W.

1996-01-01

The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project's data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup

Implementation of an integrated on-line process surveillance and diagnostic system at the Halden reactor project: MOAS

International Nuclear Information System (INIS)

Kim, I.S.; Grini, R.-E.; Nilsen, S.

2001-01-01

MOAS is an integrated on-line process surveillance and diagnostic system that uses several different models for knowledge acquisition and diagnostic reasoning, such as goal-tree success-tree model, process monitor trees, and sensor failure diagnosis trees. Within these models, the knowledge of the process and its operation, including deep knowledge, like mass balance or controller algorithm, is incorporated. During an extensive review, made as part of the integrated diagnosis system project of the Halden reactor project, MOAS (Maryland Operator Advisory System) was identified as one of the most thorough systems developed thus far. MOAS encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. The MOAS methodology was used for the NORS (Nokia Research Simulator) process at the Halden man-machine laboratory HAMMLAB of the OECD Halden reactor project. The performance tests of MOAS, implemented in G2 real-time expert system shell, show that MOAS successfully carries out its intended functions, i.e. quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The lessons learned and insights gained during the implementation and performance tests also are discussed

The MOX fuel behaviour test IFA-597.4/.5/.6/.7; Summary of in-pile fuel temperature and gas release data

Energy Technology Data Exchange (ETDEWEB)

Koike, Hisashi

2003-11-15

It is considered important to study the in-reactor behaviour of MOX fuel in order to enhance the database on such fuel. For this reason, IFA-597.4/.5/.6/.7 were included in the joint research programme of the Halden Project. The series of tests, containing two MIMAS-MOX fuel rods, both equipped with a fuel centre thermocouple and a pressure bellows transducer, has been irradiated in the Halden Reactor since July 1997 under HBWR conditions. The objectives of the test series were to study the thermal and fission gas release (FGR) behaviour of MOX fuel and to explore potential differences in behaviour between solid and hollow pellets. One of the rods had mainly solid pellets, while the other contained only hollow pellets. Both rods had an initial Pu-fissile enrichment of 6.07%. The cladding outside diameter was 9.50 mm, and the initial fuel-clad gap was 180 mum. In the course of the test, power upratings for FGR studies of the MOX fuel were planned at burnup intervals of about 10 MWd/kg MOX. The power uprating was successfully performed at approx10 MWd/kg MOX, where the estimated fuel peak temperature of the solid pellets exceeded the FGR threshold temperature for UO{sub 2} fuel, while that of the hollow pellets remained below the threshold. For the solid fuel, the temperature at onset of FGR was consistent with the empirical threshold temperature for UO{sub 2} fuel. For the hollow fuel, gas release was observed at temperatures below the threshold. FGRs at the end-of-life were approx17% for the solid pellet rod and approx14% for the hollow pellet rod, respectively. As a result of discussions in HPG meetings, IFA-597.7 was unloaded in January 2002. PIE was carried out to check in-pile pressure measurements and examine fuel structural characteristics. The discharge burn-up of the MOX fuel was 32 MWd/kg MOX as determined from in-pile power data. This report supersedes HWR-712 (June 2002) previously issued on in-pile data from IFA-597.4/5/6/7. (Author)

Analysis of pellet center temperatures measured in HALDEN IFA-224 using program FREG-3

International Nuclear Information System (INIS)

Harayama, Yasuo; Izumi, Fumio

1977-01-01

To verify the program FREG-3, we compared the calculations by FREG-3 with those by measurement in a HALDEN instrumented fuel assembly, IFA-224. FREG-3 generally gives higher pellet center temperatures than the measurement. The temperature distribution calculated by FREG-3 to estimate the stored energy in fuel rods results in safety side. (auth.)

USNRC-OECD Halden Project fuel behavior test program: experiment data report for test assemblies IFA-226 and IFA 239

International Nuclear Information System (INIS)

Laats, E.T.; MacDonald, P.E.; Quapp, W.J.

1975-12-01

The experimental data which were obtained from the IFA-226 and IFA-239 test assemblies during operation in the Halden Boiling Water Reactor are reported. Included are cladding elongation, fuel centerline temperature, internal gas pressure, and power history data from IFA-226 which were obtained from November 1971 through April 1974, and cladding elongation, diametral profile, and power history data from IFA-239 covering the period from March 1973 through April 1974. The data, presented in the form of composite graphs, have been analyzed only to the extent necessary to assure that they are reasonable and correct. A description of these mixed oxide fuel test assemblies and their instrumentation is presented. Test pin fabrication history, instrument calibration data, assembly power calibration methods, and the neutron detector data reduction technique are included as appendices

Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

International Nuclear Information System (INIS)

Hellwig, Ch.; Kasemeyer, U.

2001-01-01

In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm 3 . The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

Irradiation of inert matrix and mixed oxide fuel in the Halden test reactor

Energy Technology Data Exchange (ETDEWEB)

Hellwig, Ch.; Kasemeyer, U

2001-03-01

In a new type of fuel, called Inert Matrix Fuel (IMF), plutonium is embedded in a U-free matrix. This offers advantages for more efficient plutonium consumption, higher proliferation resistance, and for inert behaviour later in a waste repository. In the fuel type investigated at PSI, plutonium is dissolved in yttrium-stabilized zirconium oxide (YSZ), a highly radiation-resistant cubic phase, with addition of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ-based IMF is ongoing in the OECD Material Test Reactor in Halden (HBWR), together with MOX fuel (Rig IFA-651.1). The experiment is described herein and results are presented of the first 120 days of irradiation with an average assembly burnup of 47 kWd/cm{sup 3}. The results are compared with neutronic calculations performed before the experiment, and are used to model the fuel behaviour with the PSI-modified TRANSURANUS code. The measured fuel temperatures are within the expected range. An unexpectedly strong densification of the IMF during the first irradiation cycle does not alter the fuel temperatures. An explanation for this behaviour is proposed. The irradiation at higher linear heat rates during forthcoming cycles will deliver information about the fission gas release behaviour of the IMF. (author)

Analysis of selected Halden overpressure tests using the FALCON code

Energy Technology Data Exchange (ETDEWEB)

Khvostov, G., E-mail: grigori.khvostov@psi.ch [Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland); Wiesenack, W. [Institute for Energy Technology – OECD Halden Reactor Project, P.O. Box 173, N-1751 Halden (Norway)

2016-12-15

Highlights: • We analyse four Halden overpressure tests. • We determine a critical overpressure value for lift-off in a BWR fuel sample. • We show the role of bonding in over-pressurized rod behaviour. • We analytically quantify the degree of bonding via its impact on cladding elongation. • We hypothesize on an effect of circumferential cracks on thermal fuel response to overpressure. • We estimate a thermal effect of circumferential cracks based on interpretation of the data. - Abstract: Four Halden overpressure (lift-off) tests using samples with uranium dioxide fuel pre-irradiated in power reactors to a burnup of 60 MWd/kgU are analyzed. The FALCON code coupled to a mechanistic model, GRSW-A for fission gas release and gaseous-bubble swelling is used for the calculation. The advanced version of the FALCON code is shown to be applicable to best-estimate predictive analysis of overpressure tests using rods without, or weak pellet-cladding bonding, as well as scoping analysis of tests with fuels where stronger pellet-cladding bonding occurs. Significant effects of bonding and fuel cracking/relocation on the thermal and mechanical behaviour of highly over-pressurized rods are shown. The effect of bonding is particularly pronounced in the tests with the PWR samples. The present findings are basically consistent with an earlier analysis based on a direct interpretation of the experimental data. Additionally, in this paper, the specific effects are quantified based on the comparison of the data with the results of calculation. It is concluded that the identified effects are largely beyond the current traditional fuel-rod licensing analysis methods.

Research on software systems dependability at the OECD Halden Reactor Project

International Nuclear Information System (INIS)

Sivertsen, Terje; Owre, Fridtjov

2011-01-01

Two central issues related to software systems dependability are those of safety integrity and safety demonstration. A proper understanding of these two issues are important for the selection of process, methods, techniques and tools to be used in the different life cycle phases of the software. Following a brief discussion on the concept of software safety integrity and its relationship to software systems dependability, this paper gives an introduction to research problems addressed by the OECD Halden Reactor Project within this area. The paper concludes with a discussion on the important role of safety demonstration in this context. (author)

Crown Prince Regent's Resolution extending the authority of the Institute for Energy Technology regarding Kjeller and Halden

International Nuclear Information System (INIS)

1990-01-01

This Resolution extends the authority of the Institute for Energy Technology (IFE) to own and operate nuclear reactors to 31 December 1999. The Institute owns and operates the JEEP II research reactor and the Halden Boiling Water Reactor, a research reactor established as an OECD-sponsored international project. (NEA) [fr

Results of eddy current test for second round robin by Halden reactor project

International Nuclear Information System (INIS)

Iwai, Takashi; Souzawa, Shizuo; Miyata, Seiichi; Sakai, Haruyuki; Sakakura, Atsushi

1986-08-01

JMTR Hot Laboratory has executed the eddy current test of two PWR type zircaloy cladding tubes for the second round robin by Halden Reactor Project. Defects manufactured on the test specimen were revealed on a fair way to success as a function of local position, phase character and size. Influence of the fatigue crack between the two different tubes was studied through the phase angle analysis. More effort should be needed for detecting rather smaller internal defect when it was combined with external and other various type of defects. (author)

Analysis of recent fuel-disruption experiments

International Nuclear Information System (INIS)

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

1982-01-01

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

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

Science.gov (United States)

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

2018-03-04

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

An analysis of recent fuel disruption experiments

International Nuclear Information System (INIS)

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

1982-01-01

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

Post-irradiation data analysis for NRC/PNL Halden assembly IFA-431

International Nuclear Information System (INIS)

Nealley, C.; Lanning, D.D.; Cunningham, M.E.; Hann, C.R.

1979-10-01

Results are presented for the post irradiation examination performed on IFA-431, which was a 6-rod test fuel assembly irradiated in Halden Reactor, Norway, under sponsorship of the Nuclear Regulatory Commission. The irradiation conditions included: peak powers of 33 kW/m; coolant pressure and temperature of 3.3 MPa and 240 0 C, respectively; and peak burnup of 4300 MWd/MTM. IFA-431 included instrumented rods of basic boiling water reactor design, with variations in fill gas composition, gap size, and UO 2 fuel type. The irradiation was designed to measure the effect of these variations upon fuel rod thermal and mechanical performance. The post irradiation examination assessed the permanent changes to the rods, including induced radioactivity, cladding deformation, fission gas release, and fuel densification

The Halden Reactor Project workshop on improved system development using case-tools based on formal methods

International Nuclear Information System (INIS)

Gran, Bjoern Axel; Sivertsen, Terje; Stoelen, Ketil; Thunem, Harald; Zhang, Wenhui

1999-02-01

The workshop 'Improved system development using case-tools based on formal methods' was organised in Halden, December 1-2, 1998. The purpose of the workshop was to present and discuss the state-of-the-art with respect to formal approaches. The workshop had two invited presentations: 'Formality in specification and modelling: developments in software engineering practice' by John Fitzgerald (Centre for Software Reliability, UK), and 'Formal methods in industry - reaching results when correctness is not the only issue' by Oeystein Haugen (Ericsson NorARC, Norway). The workshop also had several presentations divided into three sessions on industrial experience, tools, and combined approaches. Each day there was a discussion. The first was on the effect of formalization, while the second was on the role of formal verification. At the end of the workshop, the presentations and discussions were summarised into specific recommendations. This report summarises the presentations of the speakers, the discussions, the recommendations, and the demonstrations given at the workshop (author) (ml)

Summary of the OECD Halden Reactor Project Programme on high burn-up fuel performance relevant for BWRs

International Nuclear Information System (INIS)

McGrath, M.A.

1998-01-01

The basis for the Halden Reactor Project Programme is presented together with an overview of the content of the programme for the time period 1997-1999. The concept of using both separate effects studies, to determine particular fuel properties, and integral rod behaviour studies of commercial fuel is explained. Each of the items in the programme relevant for BWRs are introduced, with most being discussed in further detail. (author)

Introduction to the Halden project and a short overview of the MMS activities at the project

International Nuclear Information System (INIS)

Owre, F.

2005-01-01

This presentation discusses the Man Machine System (MMS) research within the Halden Reactor Project located in Norway. This project is an International collaboration and the mission of this project is to improve safety at operating nuclear plants. The research activities include human reliability, knowledge management, design and evaluation of human system interfaces and control rooms, virtual reality for design, planning and training, operation and maintenance in a competitive electricity market as well as digital system safety research

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

International Nuclear Information System (INIS)

Khvostov, G.; Wiesenack, W.; Zimmermann, M.A.; Ledergerber, G.

2011-01-01

Highlights: → A model for the dynamics of axial gas redistribution in fuel rods during the LOCA is developed and coupled to the FALCON fuel behaviour code. → The first verification of the model is carried out using the data of the selected Halden LOCA tests. → According to calculation, the short rods used in the Halden tests show a small effect of the delayed gas redistribution during the clad ballooning. → The predicted effect is significant in the full length rods, eventually resulting in a considerable delay of the predicted moment of cladding rupture. → The predicted delay of cladding burst may be large enough to eventually affect the efficiency of the emergency core cooling system. - Abstract: A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code. The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from

Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

Energy Technology Data Exchange (ETDEWEB)

Aas, S.; Jamne, E.; Wullum, T.; Fjellestad, K. [Institutt for Atomenergi, OECD Halden Reactor Project, Halden (Norway)

1968-04-15

The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm{sup 2}. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO{sub 2} : the most highly exposed elements have reached 10000 MWd/t UO{sub 2}. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 {mu}C/cm{sup 3}. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel

Fuel irradiation research of Japan at Halden reactor. Achievement of cooperative researches between JAERI and several organizations in the period from 2000 to 2002 (Joint research)

International Nuclear Information System (INIS)

2004-03-01

JAERI has performed cooperative researches with several Japanese organizations utilizing the Halden Boiling Heavy Water Reactor(HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summarizes the objectives, contents and outlines of the achievements of the cooperative researches during the three years from 2000 January to 2002 December. During the period, seven cooperative researches had been carried out. Two of them had been completed and other five researches have been continued to the next three-year period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan. As the researches of fuel irradiation usually take long time for preparing test and irradiation, three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, five cooperative researches have been continued to the next three-year period. In this report, the achievements of the researches continued to the next period are not final one but a kind of progress report. (author)

Inservice inspection of Halden BWR pressure vessel

International Nuclear Information System (INIS)

Foerli, O.; Hernes, T.

1978-01-01

A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

Man-machine systems research at the OECD Halden reactor project

International Nuclear Information System (INIS)

Owre, F.; Bjorlo, T.J.; Haugset, K.

1994-01-01

The OECD Halden Reactor Project is a jointly financed research programme under the auspices of the OECD - Nuclear Energy Agency with fifteen participating countries. One of the main research topics focuses on man-machine systems. Particular attention is paid to the operator's tasks in the reactor control room environment. The overall objective of the research in this field is to provide a basis for improving today's control rooms through the introduction of computer-based solutions for the effective and safe execution of surveillance and control functions in normal as well as off-normal plant situations. The programme comprises four main activities: the verification and validation of safety critical software systems; man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; computerised operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts

Data report for the NRC/PNL Halden assembly IFA-431

International Nuclear Information System (INIS)

Hann, C.R.; Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.; Marshall, R.K.; Williford, R.E.

1978-04-01

The report presents the in-reactor data collected from the NRC/PNL Halden Assembly IFA-431 as a part of the program entitled ''Experimental Verification of Steady State Fuel Codes,'' sponsored by the Fuel Behavior Research Branch of the USNRC. The purpose of this program is to reduce the uncertainties of calculating the thermal stored energy in an operating nuclear fuel rod. The report presents fuel centerline thermocouple readings, cladding elongation monitor readings, rod internal pressure readings, and neutron detector readings. The neutron detector readings were corrected to represent rod local powers at the thermocouple locations. These data are presented in the form of plots of the variables versus time during the irradiation period from June 1975 to February 1976. Also included are descriptions of the test rationale, assembly and rod designs, test facility, instrument array and calibration, and data processing methods. Topical reports discussing specific aspects and results of the data analysis are referenced

Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

International Nuclear Information System (INIS)

Wiesenack, W.

1997-01-01

Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab

Data for FUMEX: Results from fuel behavior studies at the OECD Halden Reactor Project for model validation and development

Energy Technology Data Exchange (ETDEWEB)

Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

1997-08-01

Investigations of phenomena associated with extended or high burn-up are an important part of the fuel and materials testing programme carried out at the OECD Halden Reactor Project. The in-core studies comprise long term fuel rod behavior as well as the response to power ramps. Performance is assessed through measurements of fuel centre temperature, rod pressure, elongation of cladding and fuel stack, and cladding diameter changes obtained during full power reactor operation. Data from fuel behavior studies at the OECD Halden Reactor Project, provided for the IAEA co-ordinated research programme FUMEX, are used to elucidate short and long-term developments of fuel behavior. The examples comprise: fuel conductivity degradation manifested as a gradual temperature increase with burn-up; the influence of a combination of small gap/high fission gas release on fuel centre temperature (situation at high burn-up); fission gas release during normal operation and power ramps, and the possibility of a burn-up enhancement; PCMI reflected by cladding elongation, also for the case of a nominally open gap, and the change of interaction onset with burn-up. (author). 10 refs, 9 figs, 1 tab.

Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

International Nuclear Information System (INIS)

Horvath, M.; Munoz-Reja, C.; Tverberg, T.; Jenssen, H. K.

2010-01-01

Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP). HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as a partner in the fuel research programs. Improving the NPP operations, fuel cycles were designed to increase fuel burnup. Higher fuel burnup reduces the number of spent fuel assemblies and thus the costs of new fuel as well as the costs of back-end management. Higher burnup is reached either by prolonging the reactor cycles or by increasing the number of reactor cycles for the fuel in the core. Both ways entail additional requirements concerning fuel enrichment and burnable absorbers as additives and adjustments on the cladding material properties, such as mechanical treatment and chemical composition of the alloys. For these demands and needs ENUSA promotes the research on high burnup effects, gadolinium doped fuels and cladding material behaviour under irradiation. Various experiments, called IFA, are developed and performed also by providing materials. ENUSA collaborates with HRP on various experiments investigating the fuel densification and swelling, fission gas release, pressure limits on UO 2 and (U,Gd)O 2 fuels (IFA-504, -515, -636, -681); the cladding creep, lift-off, corrosion and hydrides on different tubing materials (IFA-567, -610, -638); instrumentation of the experiments, especially on pre-irradiated materials (IFA-533). These experiments are combined with model calculations to improve predictions for higher burnups and to maintain safety margins (IFA-515, -636, -681). Besides these unique in-pile experiments PIEs are performed as well on fuel and structural materials to complete the scope of these studies (IFA

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513. Interim report

International Nuclear Information System (INIS)

Williford, R.E.; Mohr, C.L.; Lanning, D.D.; Cunningham, M.E.; Rausch, W.N.

1980-06-01

The effects of the thermally-induced cracking and subsequent relocation of UO2 fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding axial elongations are used in a new model to solve for the effective thermal conductivity and elastic moduli of the cracked fuel column. The primary assumptions of the new model are that (1) the cracked fuel is in a hydrostatic state of stress in the (r,theta) plane, and that (2) there is no axial slipping between fuel and cladding. Three basic parameters are used to describe the cracked fuel: (1) the crack pattern, (2) the crack roughness, and (3) the fuel surface (gap) roughness. Recommendations are made on refining the model

Overview of fuel testing capabilities at the OECD Halden reactor project

Energy Technology Data Exchange (ETDEWEB)

Wiesenack, W [Institutt for Atomenergi, Halden (Norway). OECD Halden Reaktor Projekt

1994-12-31

Fuel performance and reliability investigations at the OECD Haiden Reactor Project are described. They are supported by a variety of irradiation rigs, suitable irradiation techniques and a range of instrumentation. Testing capabilities and applications are mainly aimed at exploring mechanisms of fuel behaviour and high burnup. Examples of fuel performance taken from data provided by the Halden Project for the IAEA Co-ordinated Research Programme FUMEX are presented. A number of heavily instrumented rigs to suit different test objects have been developed: base irradiation rig, gas meter rig, diameter measurement rig, ramp rig, gas flow rig, instrumented fuel assembly. In core-measurements and variety of sensors as : fuel thermocouples, bellows pressure transducers, fuel stack elongation detectors, cladding diameter gauge and cladding elongation detectors have been used. Techniques which make it possible to obtain reliable data for all relevant burnups from beginning-of-life to ultra high exposure reaching 100 Mwd/kg UO{sub 2} are described. 7 figs., 3 refs.

Fuel irradiation research of Japan at OECD Halden Reactor Project. Achievement of joint researches between JAERI and other organizations in the period from 1994 to 1996

International Nuclear Information System (INIS)

Uetsuka, Hiroshi; Nakamura, Jinichi; Kinoshita, Motoyasu

1998-01-01

JAERI has performed cooperative researches with many Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These cooperative researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1994 January to 1996 December. During the period, ten cooperative researches had been carried out, and two of them had finished during the period and other eight researches has been continued to the next three year period. There are many research items, and most of them are irradiation test researches of advanced fuel and cladding concerned with the high burnup utilization of LWR fuel or MOX fuel irradiation researches to prepare for the introduction of Plutonium utilization in LWRs. The researches of fuel irradiation usually take long time because of the characteristics of these kind of research work, and three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, eight tests have been continued after the three year period. In this report, the achievements of the continued researches to the next three year period are not final one but a kind of progress report. (author) kind of progress report. (author)

Achievements of Japanese fuel irradiation experiments in HBWR

International Nuclear Information System (INIS)

1992-10-01

OECD NEA Halden Reactor Project started in 1958, and JAERI has been participated in the Project since 1967 on behalf of Japanese Government. During the participation period, not only JAERI but also many Japanese companies and PNC, which cooperated with JAERI, have carried out many irradiation tests of fuel at HBWR. The Committee of the Halden Joint Research Programme was organized by agencies and companies, which joined the cooperative researches, and the committee has worked to promote the cooperative researches. This report summarizes the achievements of the Halden Joint Research Programme on fuel irradiation tests between Jan. 1988 and Dec. 1990., as the Halden Project renews the agreement every three years. Some researches, which have not yet been completed in the period, are also included in this report. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-15

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

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

Investigation of the resonant power oscillation in the Halden Boiling Water Reactor by autoregressive modeling

International Nuclear Information System (INIS)

Oguma, Ritsuo

1980-01-01

In the HBWR (Halden Boiling Water Reactor), there exists a resonant power oscillation with period about 0.04 Hz at power levels higher than about 9.5 MWt. While the resonant oscillation in not so large as to affect the normal reactor operation, it is significant, from the viewpoint of reactor diagnosis, to grasp its characteristics and find the cause. Noise analysis based on the autoregressive (AR) modeling technique has been made to reveal the driving source for this oscillation which led to the suggestion that it is attributed to the dynamic interference of heat exchange process between two parallel-connected steam transformers against the reactor. The present study demonstrates that the method used here is highly effective for tracing back to a noise source inducing the variation of quantities in a system, and also applicable to problems of reactor noise analysis and diagnosis. (author)

Data report for the NRC/PNL Halden Assembly IFA-432

International Nuclear Information System (INIS)

Hann, C.R.; Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.; Marshall, R.K.; Williford, R.E.

1978-08-01

The report presents the in-reactor data collected from the NRC/PNL Halden Assembly IFA-432 as a part of the program entitled ''Experimental Vertification of Steady State Fuel Codes,'' sponsored by the Fuel Behavior Research Branch of the USNRC. The purpose of this program is to reduce the uncertainties of calculating the thermal stored energy in an operating nuclear fuel rod. The report presents fuel centerline thermocouple readings, cladding elongation monitor readings, rod internal pressure readings, and neutron detector readings. The neutron detector readings were corrected to represent rod local powers at the thermocouple locations. These data are presented in the form of plots of the variables versus time during the portion of the irradiation period from December 1975 to January 1978. Also included are descriptions of the test rationale, assembly and rod designs, test facility, instrument array and calibration, and data processing methods. Topical reports discussing specific aspects and results of the data analysis are referenced. As of May 1978, the assembly burnup had reached its design goal of 20,000 MWd/MTM. However, it has been decided to leave the assembly in core to collect high burnup fission gas release data. The xenon-filled Rod 4 was replaced with the non-instrumented Rod 8 (Rod 1 design) after the first cycle. Six of the twelve original thermocouples, four of the six cladding elongation monitors, all SPND'S and all pressure transducers remain operable at this writing

The achivements of Japanese fuel irradiation experiments in HBWR

International Nuclear Information System (INIS)

Ichikawa, Michio; Yanagisawa, Kazuaki; Domoto, Kazunari

1984-02-01

OECD Halden Reactor Project celebrated the 25th anniversary in 1983. The JAERI has been participating in the Project since 1967 on behalf of Japanese Government. Since the participation, thirty-six Japanese instrumented fuel assemblies have been irradiated in HBWR. The irradiation experiments were either sponsored by JAERI or by domestic organizations under the joint research agreements with JAERI, beeing steered by the Committee for the Joint Research Programme. The cooperative efforts have attained significant contributions to the development of water reactor fuel technology in Japan. This report review the irradiation experiments of Japanese fuel assemblies. (author)

Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

International Nuclear Information System (INIS)

Backman, K.; Hallstadius, L.; Roennberg, G.

2009-01-01

Full text: There are a number of trends in the nuclear power industry, which put additional requirements on the operational flexibility and reliability of nuclear fuel, for example power uprates and longer cycles in order to increase production, higher burnup levels in order to reduce the backend cost of the fuel cycle, and lower goals for activity release from power plant operation. These additional requirements can be addressed by increasing the fuel density, improving the FG retention, improving the PCI resistance and improving the post-failure performance. In order to achieve that, Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. The additives facilitate pellet densification during sintering, enlarge the pellet grain size, and increase the creep rate. The final manufactured doped pellets reach about 0.5 % higher density within a shorter sintering time and a five times larger grain size compared with standard UO 2 fuel pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA Optima2 reloads in 2005. ADOPT pellets has been investigated in pool-side and hot cell Post Irradiation Examinations (PIEs), as well as in a ramp test and a fuel washout test in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced Fission Gas Release (FGR), improved Pellet Cladding Interaction (PCI) performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. The better FGR behaviour of ADOPT has been verified with a pool side FGR gamma measurement performed at 55 MWd/kgU, as well as transient tests in the Studsvik R2 reactor. Creep measurements performed on fresh pellets show that ADOPT has a higher creep rate which is beneficial for the PCI performance. ADOPT has also been part of a high power Halden test (IFA-677). The

Comparison with experiment of COMETHE III-L fuel rod behaviour predictions

International Nuclear Information System (INIS)

Vliet, J. van; Billaux, M.

1983-01-01

A comparison is presented between experimental results and COMETHE III-L fuel rod behaviour predictions. The first part of the paper focuses on mechanical aspects, with as main experiments, AECL X-264 and Studsvik Interramp. The second part presents the results of a wide FGR benchmarking campaign, with a reference to previous COMETHE versions. It appears that the variance between experiment and calculation has decreased by a factor four when the III-J version was improved into the III-L version. As conclusion, some COMETHE III-L calculations are presented in order to illustrate its capability of predicting fuel rod performance limits. (author)

The second eddy current testing of zircaloy tube samples from the OECD Halden reactor project at Reactor Fuel Examination Facility, Tokai, JAERI

International Nuclear Information System (INIS)

Ohwada, Isao; Nishino, Yasuharu

1986-07-01

The Reactor Fuel Examination Facility in Tokai/JAERI (Japan Atomic Energy Research Institute) joined to the second round robin programme on eddy current test of the Halden/IFE. In the programme, two zircaloy tube samples with some artificial defects were provided for measurements. To clarify the locations in axial and azimuthal directions, types and dimensions of the provided artificial defects, measured signals from eddy current test were analysed in comparison with the known defects on the calibration tube. As a result, fourteen defects were determined from the measurements. Then, the location, the type and the relative dimension of them were also revealed. The results of those eddy current test are described in this paper. (author)

A comparison of FEMAXI-III code calculations with irradiation experiments

International Nuclear Information System (INIS)

Ito, K.; Sogame, M.; Ichikawa, M.; Nakajima, T.

1981-01-01

The FEMAXI-III code calculations were compared with in-pile diameter measurements in the Halden Boiling Water Reactor, in order to check the ability to analyse the pellet-cladding mechanical interaction. The results showed generally good agreement between calculations and measurements. The Studsvik INTER-RAMP Experiments were also analysed to examine the predictability of fuel rod failures. Good agreement was obtained between calculated and measured fission gas x release. The threshold stress to cause failure was estimated by means of FEMAXI-III. (author)

Review of safety related control room function research based on experience from nuclear power plants in Finland

International Nuclear Information System (INIS)

Juslin, K.; Wahlstroem, B.; Rinttilae, E.

1985-01-01

A comprehensive human engineering research programme was established in the second half of the 1970's at the Technical Research Centre of Finland (VTT). The research is performed in cooperation with the utility companies Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (TVO) and includes topics such as Handling of alarm information, Disturbance analysis systems, Assessment of control rooms and Validation of safety parameter display systems. Reference is also made to the Finnish contribution to the OECD Halden Reactor Project (Halden) and the Nordic Liaison Committee for Atomic Energy (NKA) research projects. In this paper feasible realization alternatives of safety related control room functions are discussed on the basis of experience from the nuclear power plants in Finland, which at present are equipped with extensive process computer systems. A proposal for future power plant information systems is described. It is intended that this proposal will serve as the basis for future computer systems at nuclear power plants in Finland. (author)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

Energy Technology Data Exchange (ETDEWEB)

Slater, C.O.

1990-07-01

Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

Experience with generational changes and enhancement of competence at the OECD halden reactor project

International Nuclear Information System (INIS)

Vitanza, C.

2000-01-01

This note represents an attempt by the author to summarize his experience in managing a mid-size R and D organisation working in the nuclear sector. It basically suggests that good interpersonal relations, stable organisation structure and simple/effective routines are probably key ingredients for a successful organisation. The author is, however, not familiar with modem and sophisticated management tools and has never read a book or attended classes on management skills. Thus, this note should be taken for what it is, i.e. a simplified account of experience that does not necessarily apply to other situations and environments. (author)

Void Reactivity Effects in the Second Charge of the Halden Boiling Water Reactor; Effets Cavitaires dans la Deuxieme Charge du Reacteur a Eau Lourde Bouillante de Halden (HBWR); Ehffekty pustotnoj reaktivnosti vo vtoroj zag HBWR; Effectos de Cavitacion en la Segunda Carga del Reactor de Agua Pesada Hirviente de Halden (HBWR)

Energy Technology Data Exchange (ETDEWEB)

Lunde, J. E. [OECD Halden Reactor Project (Norway)

1964-02-15

The reactivity effect of voids caused by boiling inside the coolant channels in the second fuel charge of the Halden Boiling Heavy Water Reactor has been measured both in void-simulated zero-power experiments and under actual power conditions. The void-simulated experiments consisted of measuring the reactivity effect of introducing void columns inside thin-walled tubes to various depths. The tubes were placed at different positions between die stringers in a single 7-rod cluster element practically identical with the normal second-charge fuel elements. This experiment enables an investigation of the reactivity dependence upon void fraction, and also the reactivity dependence of steam-bubble position in the coolant channel. The experiment was carried out in the Norwegian zero-power facility NORA, with a core consisting of 36 second-charge elements and with a lattice geometry identical to the one in HBWR. The temperature dependence of the void effect was investigated in a zero-power experiment with the 100 fuel-element core of HBWR. In a single fuel element the water level inside the coolant channel was depressed to various depths, and the reactivity effect of this perturbation was measured at different temperatures in the temperature interval 50 Degree-Sign C-220 Degree-Sign C. The power void reactivity has been measured in HBWR as a function of nuclear power at different moderator temperatures between 150 Degree-Sign C and 230 Degree-Sign C at powers up to about 16 MW at the highest temperature. The power-void reactivity coefficient is an important quantity in determining the dynamic behaviour of a boiling- water reactor. The theoretical determination of this quantity is, however, complicated by the fact that knowledge about the void distribution in the core is required. The detailed power-void distribution is not easily amenable to experimental determination, and accordingly the void-simulated experiments represent a better case for testing the reactor physics

Views on quality assurance at Finnish and Swedish nuclear power plants and at Halden Reactor

International Nuclear Information System (INIS)

Hammar, L.; Lidh, B.; Wahlstroem, B.; Reiman, T.

2001-06-01

The paper reports on a study within the Nordic Nuclear Safety Research, NKS on quality systems at nuclear installations in Finland, Norway and Sweden. In the study a total of 74 people at the NPPs in Barsebaeck, Forsmark, Loviisa, Olkiluoto, Oskarshamn and Ringhals, and at the research reactor in Halden were interviewed in the period 30 August to 13 December 2000 concerning their views in regard of quality and quality systems. The study was concluded with a seminar held in the Ringhals nuclear power plant in Januar 2001. The study covered a number of aspects in regard of quality management, including the quality concept, quality systems, topical quality issues and approaches, rules and procedures, competency and training, the process approach to quality management, the promotion of quality consciousness and future prospects. The study reflects the significant progress made in the management of quality in nuclear power in the Nordic countries since the early phase in the seventies. The most distinctive characteristic of today's approach to quality is seen in that responsibility for the quality is assumed directly in conjunction with the working processes. It could be noted that the work patterns at the nuclear installations have been largely modified during the recent years as a result of persistent endeavours to continuously improve the quality of operation. Challenges were seen in currently reduced revenues due to descending electricity prices and the likely prospect of further increased regulatory safety requirements. The report is aimed for those working with quality issues at the nuclear power plants as well as for those interested in quality management in general or in the safety aspects of nuclear power in particular. (au)

Forest gene conservation from the perspective of the international community

Science.gov (United States)

M. Hosny El-Lakany

2017-01-01

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

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

Science.gov (United States)

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

2013-04-01

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

Human-machine interface aspects and use of computer-based operator support systems in control room upgrades and new control room designs for nuclear power plants

International Nuclear Information System (INIS)

Berg, O.

1997-01-01

At the Halden Project efforts are made to explore the possibilities through design, development and validation of Computer-based Operator Support Systems (COSSes) which can assist the operators in different operational situations, ranging from normal operation to disturbance and accident conditions. The programme comprises four main activities: 1) verification and validation of safety critical software systems; 2) man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; 3) computerized operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and 4) control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts. The paper presents the status of this development programme, including descriptions of specific operator support functions implemented in the simulator-based, experimental control room at Halden (HAMMLAB, HAlden Man-Machine LABoratory). These operator aids comprise advanced alarms systems, diagnostic support functions, electronic procedures, critical safety functions surveillance and accident management support systems. The different operator support systems development at the Halden Project are tested and evaluated in HAMMLAB with operators from the Halden Reactor, and occasionally from commercial NPPs, as test subjects. These evaluations provide data on the merits of different operator support systems in an advanced control room setting, as well as on how such systems should be integrated to enhance operator performance. The paper discusses these aspects and the role of computerized operator support systems in plant operation based on the experience from this work at the Halden Project. 15 refs, 5 figs

Human-machine interface aspects and use of computer-based operator support systems in control room upgrades and new control room designs for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Berg, O [Institutt for Energiteknikk, OECD Halden Reactor Project (Netherlands)

1997-07-01

At the Halden Project efforts are made to explore the possibilities through design, development and validation of Computer-based Operator Support Systems (COSSes) which can assist the operators in different operational situations, ranging from normal operation to disturbance and accident conditions. The programme comprises four main activities: 1) verification and validation of safety critical software systems; 2) man-machine interaction research emphasizing improvements in man-machine interfaces on the basis of human factors studies; 3) computerized operator support systems assisting the operator in fault detection/diagnosis and planning of control actions; and 4) control room development providing a basis for retrofitting of existing control rooms and for the design of advanced concepts. The paper presents the status of this development programme, including descriptions of specific operator support functions implemented in the simulator-based, experimental control room at Halden (HAMMLAB, HAlden Man-Machine LABoratory). These operator aids comprise advanced alarms systems, diagnostic support functions, electronic procedures, critical safety functions surveillance and accident management support systems. The different operator support systems development at the Halden Project are tested and evaluated in HAMMLAB with operators from the Halden Reactor, and occasionally from commercial NPPs, as test subjects. These evaluations provide data on the merits of different operator support systems in an advanced control room setting, as well as on how such systems should be integrated to enhance operator performance. The paper discusses these aspects and the role of computerized operator support systems in plant operation based on the experience from this work at the Halden Project. 15 refs, 5 figs.

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

Science.gov (United States)

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

2005-05-01

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

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

Science.gov (United States)

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

2018-01-01

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

Fission gas release of MOX with heterogeneous structure

International Nuclear Information System (INIS)

Nakae, N.; Akiyama, H.; Kamimura, K; Delville, R.; Jutier, F.; Verwerft, M.; Miura, H.; Baba, T.

2015-01-01

It is very useful for fuel integrity evaluation to accumulate knowledge base on fuel behavior of uranium and plutonium mixed oxide (MOX) fuel used in light water reactors (LWRs). Fission gas release is one of fuel behaviors which have an impact on fuel integrity evaluation. Fission gas release behavior of MOX fuels having heterogeneous structure is focused in this study. MOX fuel rods with a heterogeneous fuel microstructure were irradiated in Halden reactor (IFA-702) and the BR-3/BR-2 CALLISTO Loop (CHIPS program). The 85 Kr gamma spectrometry measurements were carried out in specific cycles in order to examine the concerned LHR (Linear Heat Rate) for fission gas release in the CHIPS program. The concerned LHR is defined in this paper to be the LHR at which a certain additional fission gas release thermally occurs. Post-irradiation examination was performed to understand the fission gas release behavior in connection with the pellet microstructure. The followings conclusions can be made from this study. First, the concerned LHR for fission gas release is estimated to be in the range of 20-23 kW/m with burnup over 37 GWd/tM. It is moreover guessed that the concerned LHR for fission gas release tends to decrease with increasing burnup. Secondly It is observed that FGR (fission gas release rate) is positively correlated with LHR when the LHR exceeds the concerned value. Thirdly, when burnup dependence of fission gas release is discussed, effective burnup should be taken into account. The effective burnup is defined as the burnup at which the LHR should be exceed the concerned value at the last time during all the irradiation period. And fourthly, it appears that FGR inside Pu spots is higher than outside and that retained (not released) fission gases mainly exist in the fission gas bubbles. Since fission gases in bubbles are considered to be easily released during fuel temperature increase, this information is very important to estimate fission gas release behavior

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

Institute of Scientific and Technical Information of China (English)

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

2008-01-01

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

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

Science.gov (United States)

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

2016-11-01

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

Studies of Nuclear Fuel Performance Using On-site Gamma-ray Spectroscopy and In-pile Measurements

International Nuclear Information System (INIS)

Matsson, Ingvar

2006-01-01

Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebaeck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA

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

NARCIS (Netherlands)

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

2013-01-01

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

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

Science.gov (United States)

Bennett, Kochise; Kowalewski, Markus; Mukamel, Shaul

2016-02-09

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

Comparison of FISGAS swelling and gas release predictions with experiment

International Nuclear Information System (INIS)

Ostensen, R.W.

1979-01-01

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

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

NARCIS (Netherlands)

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

2017-01-01

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

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

NARCIS (Netherlands)

Cohen, Emily

2017-01-01

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

Fourth session: perspectives and internationalization of nuclear research

International Nuclear Information System (INIS)

Bugat, S.; Girardin, G.; Vitanza, C.

2005-01-01

The purpose of the atomic simulation is to deduce the behaviour of irradiated materials from the effects of irradiation at the atomic scale that are well understood. The main difficulties and recent breakthroughs concerning the simulation of the primary damage and the microstructure due to irradiation and of the hardening effect of irradiation are reviewed. It is shown that simulation tools are far to be able to replace real irradiation experiments but their maturity is so high that they will allow us to optimize the design and operations of irradiation experiments in a near future. The second article is dedicated to the Norwegian Halden research reactor that was at the very beginning of its operating life (1958) an irradiation facility broadly open to the international nuclear community. The Halden reactor is a boiling reactor, cooled and moderated with heavy water (14 m 3 ) and whose thermal power output is 20 MW. The steam generated (30 tons/h) is used to operate a paper mill. 12 experimental loops with in-core test rigs are available. In 1999 about 68% of the studies performed at Halden was dedicated to high burnup fuels and 32% to materials. (A.C.)

Fourth session: perspectives and internationalization of nuclear research; Session 4: Les perspectives, l'internationalisation de la R and D

Energy Technology Data Exchange (ETDEWEB)

Bugat, S. [Electricite de France, 77 - Moret sur Loing (France); Girardin, G. [AREVA-FRAMATOME-ANP, Centre Technique, NFTC, 71 - Le Creusot (France); Vitanza, C. [Organisation for Economic Co-Operation and Development, Nuclear Energy Agency (OECD/NEA), 75 - Paris (France)

2005-07-01

The purpose of the atomic simulation is to deduce the behaviour of irradiated materials from the effects of irradiation at the atomic scale that are well understood. The main difficulties and recent breakthroughs concerning the simulation of the primary damage and the microstructure due to irradiation and of the hardening effect of irradiation are reviewed. It is shown that simulation tools are far to be able to replace real irradiation experiments but their maturity is so high that they will allow us to optimize the design and operations of irradiation experiments in a near future. The second article is dedicated to the Norwegian Halden research reactor that was at the very beginning of its operating life (1958) an irradiation facility broadly open to the international nuclear community. The Halden reactor is a boiling reactor, cooled and moderated with heavy water (14 m{sup 3}) and whose thermal power output is 20 MW. The steam generated (30 tons/h) is used to operate a paper mill. 12 experimental loops with in-core test rigs are available. In 1999 about 68% of the studies performed at Halden was dedicated to high burnup fuels and 32% to materials. (A.C.)

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

Science.gov (United States)

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

2018-03-01

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

Fuel Element Experience at the Halden Boiling Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Aas, S. [OECD Halden Reactor Project, Halden (Norway); Videm, K.; Hanevik, A. [Institutt for Atomenergi, Kjeller (Norway)

1968-04-15

The penalty for neutron absorbing materials is higher for a reactor moderated with heavy water than one with light water. As Zircaloy and enriched uranium were not readily available in 1954 when the design of the first fuel charge for HBWR was frozen, fuel elements of natural uranium metal clad in a specially developed aluminium alloy (A 1 0.3% Fe, 0.03% Si) were used. The temperature was limited to 150 Degree-Sign C and with this limitation the general behaviour of the elements was good. In I960, in another effort to maintain a good neutron economy, a couple of elements with as thin cladding as 0.25 mm A1S1 316, stainless steel with an unsegmented length of 2 m supported by wire grid spacers were tested. These elements with 1.5% enriched UO{sub 2} behaved satisfactorily at 150'C. Elements of a rather similar construction failed due to stress corrosion during the later operation at 230 'C. The reason for the different behaviour is probably the higher stresses in the cladding, due to the increased pressure, possibly combined with a short period with a high chloride content in the heavy water. The second fuel core with 1.5% enriched UO{sub 2} clad in Zircaloy-2 was installed in order to permit an increase in temperature to 230 Degree-Sign C and in power from 5 to 20 MW(th). The maximum burnup obtained is 11000 MWd/t and the maximum heat rating 375 W/cm with no fracture failure and practically no change in appearance according to the post-irradiation examination. One element was deliberately taken to burn-out conditions by throttling the water flow. After a series of burn-outs, the element finally failed because of over-temperature. The successful use of aluminium cladding at 150 Degree-Sign C mitiated an effort for making aluminium alloys suitable for normal power reactor operation. Promising properties were found for an alloy (designated IFA 3 aluminium) with A1 10% Si, 1% Ni, 1% Mg, 0.3% Fe + Ti. Despite increase in corrosion rate under heat transfer conditions, aluminium-clad elements have reached a bumup of 9000 MWd/t (av.) at a heat flux as high as 150 W/cm{sup 2}. (author)

Trisomy 9 Mosaicism Diagnosed In Utero

Directory of Open Access Journals (Sweden)

Hironori Takahashi

2010-01-01

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

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

Science.gov (United States)

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

2009-10-01

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

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

Directory of Open Access Journals (Sweden)

Jacques Beltrand

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

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

Science.gov (United States)

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

2015-04-01

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

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

Science.gov (United States)

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

2016-03-01

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

Intrauterine Intervention for the Treatment of Fetal Growth Restriction.

Science.gov (United States)

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

2016-01-01

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

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

Directory of Open Access Journals (Sweden)

Kurahashi Hiroki

2011-08-01

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

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

Science.gov (United States)

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

2017-07-01

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

Workshop meeting

International Nuclear Information System (INIS)

Veland, Oeystein

2004-04-01

1-2 September 2003 the Halden Project arranged a workshop on 'Innovative Human-System Interfaces and their Evaluation'. This topic is new in the HRP 2003-2005 programme, and it is important to get feedback from member organizations to the work that is being performed in Halden. It is also essential that relevant activities and experiences in this area from the member organizations are shared with the Halden staff and other HRP members. Altogether 25 persons attended the workshop. The workshop had a mixture of presentations and discussions, and was chaired by Dominique Pirus of EDF, France. Day one focused on the HRP/IFE activities on Human-System Interface design, including Function-oriented displays, Ecological Interface Design, Task-oriented displays, as well as work on innovative display solutions for the oil and gas domain. There were also presentations of relevant work in France, Japan and the Czech Republic. The main focus of day two was the verification and validation of human-system interfaces, with presentations of work at HRP on Human-Centered Validation, Criteria-Based System Validation, and Control Room Verification and Validation. The chairman concluded that it was a successful workshop, although one could have had more time for discussions. The Halden Project got valuable feedback and viewpoints on this new topic during the workshop, and will consider all recommendations related to the future work in this area. (Author)

Analysis of Operators Comments on the PSF Questionnaire of the Task Complexity Experiment 2003/2004

Energy Technology Data Exchange (ETDEWEB)

Torralba, B.; Martinez-Arias, R.

2007-07-01

Human Reliability Analysis (HRA) methods usually take into account the effect of Performance Shaping Factors (PSF). Therefore, the adequate treatment of PSFs in HRA of Probabilistic Safety Assessment (PSA) models has a crucial importance. There is an important need for collecting PSF data based on simulator experiments. During the task complexity experiment 2003-2004, carried out in the BWR simulator of Halden Man-Machine Laboratory (HAMMLAB), there was a data collection on PSF by means of a PSF Questionnaire. Seven crews (composed of shift supervisor, reactor operator and turbine operator) from Swedish Nuclear Power Plants participated in the experiment. The PSF Questionnaire collected data on the factors: procedures, training and experience, indications, controls, team management, team communication, individual work practice, available time for the tasks, number of tasks or information load, masking and seriousness. The main statistical significant results are presented on Performance Shaping Factors data collection and analysis of the task complexity experiment 2003/2004 (HWR-810). The analysis of the comments about PSFs, which were provided by operators on the PSF Questionnaire, is described. It has been summarised the comments provided for each PSF on the scenarios, using a content analysis technique. (Author)

Analysis of Operators Comments on the PSF Questionnaire of the Task Complexity Experiment 2003/2004

International Nuclear Information System (INIS)

Torralba, B.; Martinez-Arias, R.

2007-01-01

Human Reliability Analysis (HRA) methods usually take into account the effect of Performance Shaping Factors (PSF). Therefore, the adequate treatment of PSFs in HRA of Probabilistic Safety Assessment (PSA) models has a crucial importance. There is an important need for collecting PSF data based on simulator experiments. During the task complexity experiment 2003-2004, carried out in the BWR simulator of Halden Man-Machine Laboratory (HAMMLAB), there was a data collection on PSF by means of a PSF Questionnaire. Seven crews (composed of shift supervisor, reactor operator and turbine operator) from Swedish Nuclear Power Plants participated in the experiment. The PSF Questionnaire collected data on the factors: procedures, training and experience, indications, controls, team management, team communication, individual work practice, available time for the tasks, number of tasks or information load, masking and seriousness. The main statistical significant results are presented on Performance Shaping Factors data collection and analysis of the task complexity experiment 2003/2004 (HWR-810). The analysis of the comments about PSFs, which were provided by operators on the PSF Questionnaire, is described. It has been summarised the comments provided for each PSF on the scenarios, using a content analysis technique. (Author)

IFPE/IFA-597.3, centre-line temperature, fission gas release and clad elongation at high burn-up (60-62 MWd/kg)

International Nuclear Information System (INIS)

Turnbull, J.A.

2003-01-01

Description: The fuel segments for the high burn-up integral rod behaviour test IFA-597 were taken from fuel rod 33-25065, which was irradiated in the Ringhals 1 BWR for approximately 12 years. The irradiation of this rod and its sibling rod 33-25046 was performed in two stages. During the first irradiation, 1980 to 1986, the rods were part of Ringhals assembly 6477 and an approximate rod averaged burn-up of 31 MWd/kg UO 2 was reached. The rods were then placed into fuel assembly 9902 for a second period of irradiation from 1986 to 1992. The location of the fuel rods 33-25065 and 33-25046 in this assembly were in positions 9902/D and 9902/E4 respectively. A final rod averaged burn-up of 52 MWd/kg UO 2 was achieved. The burn-up at the location of the Halden segments was estimated as 59 MWd/kg UO 2 , well beyond the formation of High Burn-up Structure (Hobs) formation at the pellet rim. At the rim, the burn-up was estimated as 130 MWd/kg UO 2 . After commercial irradiation, PIE was performed at Studsvik. Inner and outer clad oxide thickness measurements were 42 and 5 microns respectively. The measured cold rod diameter varied between 12.20 and 12.25 mm, thus only a small amount of creep-down had occurred from the original diameter of 12.25 mm. Cold gap measurements were taken by diametral compression of the clad onto the fuel. The stiffness changes twice during these measurements, the first (relocated gap) associated with the onset of pellet fragment movement, the second (compressed gap) when the fragments are together and the pellet is compressed. For these rods, the compressed diametral gap was measured as 30 microns. This is in agreement with the pellet and cladding being in contact during the final irradiation cycle, i.e., at ∼12 kW/m. FGR measurements were made after puncturing and values of 2.5%-3.3% were calculated from the extracted gas. The uncertainty is due to different methods of calculation. Ceramography showed a normal crack pattern and no evidence of

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

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

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

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

Science.gov (United States)

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

2017-12-01

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

The MOX Fuel Behaviour Test IFA-597.4: Temperature And Pressure Data To A Burn-Up Of 5.4 MWd/kg MOX

International Nuclear Information System (INIS)

McGrath, M. A.; Teshima, H.

1998-02-01

Characterising the behaviour of MOX fuel is becoming increasingly important as many commercial reactors are or will be operating with this type of fuel. With this as a driving force, a new joint programme experiment, IFA-597.4, has been loaded into the reactor at Halden for the purpose of establishing the fission gas release behaviour of MOX fuel. Both annular and solid pellet fuel is being utilised and the irradiation is being conducted such that the fuel is initially operated below the onset of fission gas release. The fuel will later be subjected to small power up ratings which will be held for short periods of time. These are designed to bring the fuel to just above the temperature threshold for fission gas release thus allowing the FGR behaviour of both solid and annular MOX fuel to be established. The rig contains two fuel rods of active length 220 mm and diameter 8.05 mm. Both fuel rods contain MOX fuel with an initial Pu-fissile content of 6.07% and both are instrumented with a fuel centre thermocouple and a pressure transducer. The test is being performed under HBWR conditions and at the time of the reactor shutdown at the end of 1997 a mean burn-up of 5.4 MWd/kg MOX had been achieved with the rods at an average rating of 30 kW/m. The rod pressure data show that no fission gas had been released up to the shutdown. The fuel centre temperatures of both rods exhibit an initial increase concurrent with a fall in the monitored rod internal pressures as a result of fuel densification. It was estimated that about 1-1.4% fuel densification by volume had occurred in the two rods by a burn-up of about 3 MWd/kg MOX. (author)

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

Science.gov (United States)

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

2014-08-01

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

Prediction of adverse pregnancy outcomes using uterine artery Doppler imaging at 22-24 weeks of pregnancy: A North Indian experience

Directory of Open Access Journals (Sweden)

Deepti Verma

2016-06-01

Full Text Available Objective: The aim of this study was to assess the predictive value of uterine artery Doppler imaging at 22-24 weeks of gestation for adverse pregnancy outcomes. Materials and Methods: This was a prospective study in which uterine artery Doppler was performed at 22-24 weeks of gestation in 165 pregnant women with singleton pregnancies. A pulsatility index (PI more than 1.45 or bilateral uterine notching was labeled as abnormal Doppler. The pregnancy outcome was assessed in terms of normal outcome, preeclampsia, fetal growth restriction (FGR, low birth weight, spontaneous preterm delivery, oligohydramnios, fetal loss or at least one adverse outcome. Results: Out of 165 patients, 35 (21.2% had abnormal second trimester uterine artery Doppler. In pregnancies that resulted in preeclampsia (PE, (n=21, FGR, (n=21, and low birth weight (n=39, the median uterine artery PI was higher (1.52, 1.41, and 1.27 respectively. In the presence of abnormal Doppler, the risk of PE [OR=10.7, 95% confidence interval (CI: (3.91-29.1; p<0.001], FGR [OR=4.34, 95% CI: (1.62-11.6; p=0.002], low birth weight [OR=6.39, 95% CI: (3.16-12.9; p<0.001] and the risk of at least one obstetric complication [OR=8.73, 95% CI: (3.5-21.3; p<0.001] was significantly high. The positive predictive value of abnormal uterine artery Doppler was highest for preeclampsia (36.84% among all adverse pregnancy outcomes assessed. Conclusion: Uterine artery Doppler ultrasonography at 22-24 weeks of gestation is a significant predictor of at least one adverse pregnancy outcome, with the highest prediction for preeclampsia.

Extracellular peptidases of the cereal pathogen Fusarium graminearum.

Directory of Open Access Journals (Sweden)

Rohan George Thomas Lowe

2015-11-01

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

Comparison of the ENIGMA code with experimental data on thermal performance, stable fission gas and iodine release at high burnup

Energy Technology Data Exchange (ETDEWEB)

Killeen, J C [Nuclear Electric plc, Barnwood (United Kingdom)

1997-08-01

The predictions of the ENIGMA code have been compared with data from high burn-up fuel experiments from the Halden and RISO reactors. The experiments modelled were IFA-504 and IFA-558 from Halden and the test II-5 from the RISO power burnup test series. The code has well modelled the fuel thermal performance and has provided a good measure of iodine release from pre-interlinked fuel. After interlinkage the iodine predictions remain a good fit for one experiment, but there is significant overprediction for a second experiment (IFA-558). Stable fission gas release is also well modelled and the predictions are within the expected uncertainly band throughout the burn-up range. This report presents code predictions for stable fission gas release to 40GWd/tU, iodine release measurements to 50GWd/tU and thermal performance (fuel centre temperature) to 55GWd/tU. Fuel ratings of up to 38kW/m were modelled at the high burn-up levels. The code is shown to accurately or conservatively predict all these parameters. (author). 1 ref., 6 figs.

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

Directory of Open Access Journals (Sweden)

Borna Sedigheh

2009-10-01

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

Evaluation of placenta in foetal demise and foetal growth restriction.

Science.gov (United States)

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

2013-11-01

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

Annual report 1988

International Nuclear Information System (INIS)

1989-01-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. Nuclear power research at the Institute is performed within the international OECD Halden Reactor Project. The nuclear reactor fuel experiments conducted in 1988 included investigations of the corrosion of fuel cladding materials and measurements of gas pressures inside fuel rods. The effect of fuel pellets roughness on heat transfer to the fuel cladding, and the use of different filler gases in the fuel rods were also studied. Work continiued on the development of a new thermometer to measure fuel temperatures and on other instruments. The second major research function of the Halden Project involves the development of computerized reactor process controls. This represents over one-third of the total research programme and emphasizes analysis of human performance and computer-based control and monitoring of the operational aspects of nuclear power plants. This work has resulted in the development of a monitoring system to help operators diagnose and mitigate the consequences of an accident. In addition, the SCORPIO Core Surveillance System, which was developed in Halden, was installed in the control room of a Swedish reactor at Ringhals in 1988

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

Science.gov (United States)

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

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

Mark Robert Dilworth

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

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

Science.gov (United States)

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

2018-02-01

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

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

Science.gov (United States)

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

2016-10-28

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

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

NARCIS (Netherlands)

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

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

Measurement station for interim inspections of Lightbridge metallic fuel rods at the Halden Boiling Water Reactor

Science.gov (United States)

Hartmann, C.; Totemeier, A.; Holcombe, S.; Liverud, J.; Limi, M.; Hansen, J. E.; Navestad, E. AB(; )

2018-01-01

Lightbridge Corporation has developed a new Uranium-Zirconium based metallic fuel. The fuel rods aremanufactured via a co-extrusion process, and are characterized by their multi-lobed (cruciform-shaped) cross section. The fuel rods are also helically-twisted in the axial direction. Two experimental fuel assemblies, each containing four Lightbridge fuel rods, are scheduled to be irradiated in the Halden Boiling Water Reactor (HBWR) starting in 2018. In addition to on-line monitoring of fuel rod elongation and critical assembly conditions (e.g. power, flow rates, coolant temperatures, etc.) during the irradiation, several key parameters of the fuel will be measured out-of-core during interim inspections. An inspection measurement station for use in the irradiated fuel handling compartment at the HBWR has therefore been developed for this purpose. The multi-lobed cladding cross section combined with the spiral shape of the Lightbridge metallic fuel rods requires a high-precision guiding system to ensure good position repeatability combined with low-friction guiding. The measurement station is equipped with a combination of instruments and equipment supplied from third-party vendors and instruments and equipment developed at Institute for Energy Technology (IFE). Two sets of floating linear voltage differential transformer (LVDT) pairs are used to measure swelling and diameter changes between the lobes and the valleys over the length of the fuel rods. Eddy current probes are used to measure the thickness of oxide layers in the valleys and on the lobe tips and also to detect possible surface cracks/pores. The measurement station also accommodates gamma scans. Additionally, an eddy-current probe has been developed at IFE specifically to detect potential gaps or discontinuities in the bonding layer between the metallic fuel and the Zirconium alloy cladding. Potential gaps in the bonding layer will be hidden behind a 0.5-1.0 mm thick cladding wall. It has therefore been

Thermal behaviour of high burnup PWR fuel under different fill gas conditions

International Nuclear Information System (INIS)

Tverberg, T.

2001-01-01

During its more than 40 years of existence, a large number of experiments have been carried out at the Halden Reactor Project focusing on different aspects related to nuclear reactor fuel. During recent years, the fuels testing program has mainly been focusing on aspects related to high burnup, in particular in terms of fuel thermal performance and fission gas release, and often involving reinstrumentation of commercially irradiated fuel. The paper describes such an experiment where a PWR rod, previously irradiated in a commercial reactor to a burnup of ∼50 MWd/kgUO 2 , was reinstrumented with a fuel central oxide thermocouple and a cladding extensometer together with a high pressure gas flow line, allowing for different fill gas compositions and pressures to be applied. The paper focuses on the thermal behaviour of such LWR rods with emphasis on how different fill gas conditions influence the fuel temperatures and gap conductance. Rod growth rate was also monitored during the irradiation in the Halden reactor. (author)

An Investigation of an Irradiated Fuel Pin by Measurement of the Production of Fast Neutrons in a Thermal Column and by Pile Oscillation Technique

International Nuclear Information System (INIS)

Gustavsson, Veine

1968-05-01

A fuel pin irradiated to about 3400 MWd/tU from the Halden reactor has been investigated by a measurement of the production of fast neutrons in a thermal column and by pile oscillator technique in the central channel of the reactor R1. Calibration was made by using samples with different U 235 enrichments. The thermal column experiment gives the quantity ave(νΣ f ) (average in the thermal column spectrum) for the Halden sample. Σ f is the macroscopic fission cross section and ν is the number of fast neutrons produced per fission. The result of the oscillator measurements is a value of ave(Σ a ) - w ave(Σ f ) (average in the central channel spectrum) for the irradiated sample, w is the importance of a fast neutron relative to a thermal one and ave(Σ a ) is the macroscopic absorption cross section. The results from both the experiments have been compared with values calculated by the REBUS code and the agreement was good

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

NARCIS (Netherlands)

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

2018-01-01

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

Dynamic conservation of forest genetic resources in 33 European countries

NARCIS (Netherlands)

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

2013-01-01

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

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

NARCIS (Netherlands)

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

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

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

NARCIS (Netherlands)

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

2016-01-01

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

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

Science.gov (United States)

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

2018-04-01

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

Pramana – Journal of Physics | Indian Academy of Sciences

Indian Academy of Sciences (India)

BARC has developed large-area silicon detectors in collaboration with BEL to be used in the pre-shower detector of the CMS experiment at CERN. The use of floating guard rings (FGR) in improving breakdown voltage and reducing leakage current of silicon detectors is well-known. In the present work, it has been ...

On-line monitoring for calibration reduction

International Nuclear Information System (INIS)

Hoffmann, Mario; Gran, Frauke Schmitt; Thunem, Harald P-J.

2004-04-01

On-Line Monitoring (OLM) of a channel's calibration state evaluates instrument channel performance by assessing its consistency with other plant indications. Industry and experience at several plants has shown this overall approach to be very effective in identifying instrument channels that are exhibiting degrading or inconsistent performance characteristics. The Halden Reactor Project has developed the signal validation system PEANO, which can be used to assist with the tasks of OLM. To further enhance the PEANO System for use as a calibration reduction tool, the following two additional modules have been developed; HRP Prox, which performs pre-processing and statistical analysis of signal data, Batch Monitoring Module (BMM), which is an off-line batch monitoring and reporting suite. The purpose and functionality of the HRP Prox and BMM modules are discussed in this report, as well as the improvements made to the PEANO Server to support these new modules. The Halden Reactor Project has established a Halden On-Line Monitoring User Group (HOLMUG), devoted to the discussion and implementation of on-line monitoring techniques in power plants. It is formed by utilities, vendors, regulatory bodies and research institutes that meet regularly to discuss implementation aspects of on-line monitoring, technical specification changes, cost-benefit analysis and regulatory issues. (Author)

Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

International Nuclear Information System (INIS)

Casal, V.; Malang, S.; Rust, K.

1982-10-01

For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

Lessons learned on digital systems safety

International Nuclear Information System (INIS)

Sivertsen, Terje

2005-06-01

A decade ago, in 1994, lessons learned from Halden research activities on digital systems safety were summarized in the reports HWR-374 and HWR-375, under the title 'A Lessons Learned Report on Software Dependability'. The reports reviewed all activities made at the Halden Project in this field since 1977. As such, the reports provide a wealth of information on Halden research. At the same time, the lessons learned from the different activities are made more accessible to the reader by being summarized in terms of results, conclusions and recommendations. The present report provides a new lessons learned report, covering the Halden Project research activities in this area from 1994 to medio 2005. As before, the emphasis is on the results, conclusions and recommendations made from these activities, in particular how they can be utilized by different types of organisations, such as licensing authorities, safety assessors, power companies, and software developers. The contents of the report have been edited on the basis of input from a large number of Halden work reports, involving many different authors. Brief summaries of these reports are included in the last part of the report. (Author)

Studying the effects of operators' problem solving behaviour when using a diagnostic expert system developed for the nuclear industry

International Nuclear Information System (INIS)

Holmstroem, C.B.O.; Volden, F.S.; Endestad, T.

1992-01-01

This paper describes an experiment with the purpose to also illustrate and discuss some of the methodological problems when empirically studying problem solving. The experiment which was the second in a series, conducted at the OECD Halden Reactor Project, aimed to assess the effect on nuclear power plant operators diagnostic behaviour when using a rule-based diagnostic expert system. The rule-based expert system used in the experiment is called DISKET (Diagnosis System Using Knowledge Engineering Technique) and was originally developed by the Japan Atomic Energy Research Institute (JAERI). The experiment was performed in the Halden man-machine laboratory using a full scope pressurized water reactor simulator. Existing data collection methods and experimental design principles includes possibilities but also limitations. This is discussed and experiences are presented. Operator performance in terms of quality of diagnosis is improved by the use of DISKET. The use of the DISKET system also influences operators problem solving behaviour. The main difference between the two experimental conditions can be characterized as while the DISKET users during the diagnosis process are following a strategy which is direct and narrowed, the non-DISKET users are using a much broader and less focused search when trying to diagnose a disturbance. (author)

Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

Energy Technology Data Exchange (ETDEWEB)

Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

2013-03-15

The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

An EG-VEGF-dependent decrease in homeobox gene NKX3.1 contributes to cytotrophoblast dysfunction: a possible mechanism in human fetal growth restriction.

Science.gov (United States)

Murthi, P; Brouillet, S; Pratt, A; Borg, Aj; Kalionis, B; Goffin, F; Tsatsaris, V; Munaut, C; Feige, Jj; Benharouga, M; Fournier, T; Alfaidy, N

2015-07-21

Idiopathic fetal growth restriction (FGR) is frequently associated with placental insufficiency. Previous reports have provided evidence that EG-VEGF (endocrine gland derived-vascular endothelial growth factor), a placental secreted protein, is expressed during the first trimester of pregnancy, controls both trophoblast proliferation and invasion, and its increased expression is associated with human FGR. In this study, we hypothesise that EG-VEGF-dependent change in placental homeobox gene expressions contribute to trophoblast dysfunction in idiopathic FGR. The changes in EG-VEGF-dependent homeobox gene expressions were determined using a Homeobox gene cDNA array on placental explants of 8-12 weeks' gestation after stimulation with EG-VEGF in vitro for 24 hours. The Homeobox gene array identified a >5-fold increase in HOXA9, HOXC8, HOXC10, HOXD1, HOXD8, HOXD9 and HOXD11, while NKX 3.1 showed a >2 fold-decrease in mRNA expression compared to untreated controls. Homeobox gene NKX3.1 was selected as a candidate because it is a downstream target of EG-VEGF and its expression and functional role are largely unknown in control and idiopathic FGR-affected placentae. Real-time PCR and immunoblotting showed a significant decrease in NKX3.1 mRNA and protein levels, respectively, in placentae from FGR compared to control pregnancies. Gene inactivation in vitro using short-interference RNA specific for NKX3.1 demonstrated an increase in BeWo cell differentiation and a decrease in HTR8-SVneo proliferation. We conclude that the decreased expression of homeobox gene NKX3.1 down-stream of EG-VEGF may contribute to the trophoblast dysfunction associated with idiopathic FGR pregnancies.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-01

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

Integrated System Validation Usability Questionnaire: Information Display Element; Desarrollo del Cuestionario de Facilidad de Uso para la Validación de Sistemas Integrados: Displays de Información

Energy Technology Data Exchange (ETDEWEB)

Garcés, Ma. I.; Torralba, B.

2015-07-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to information displays, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed and the models proposed by the nuclear energy industry and their technical support organizations, mainly, the United States Electric Power Research Institute, EPRI, are analyzed. In the remaining stages, general and specific guidelines for information displays, in particular, display pages, formats, elements and data quality and update rate recommendations are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMMLAB.

Integrated System Validation Usability Questionnaire: Computerized Procedures; Desarrollo del Cuestionario de Facilidad de Uso para la Validación de Sistemas Integrados: Procedimientos de Operacion

Energy Technology Data Exchange (ETDEWEB)

Garcés, Ma. I.; Torralba, B.

2015-07-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to operating procedures, both computer and paper-based, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed, the models used by the nuclear industry and their technical support organizations, mainly, the Electric Power Research Institute, EPRI, are analyzed, and scientist advances are also explored. In the remaining stages, general and specific guidelines for computerized and paper-based procedures are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMLAB.

Integrated System Validation Usability Questionnaire: Information Display Element

International Nuclear Information System (INIS)

Garcés, Ma. I.; Torralba, B.

2015-01-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to information displays, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed and the models proposed by the nuclear energy industry and their technical support organizations, mainly, the United States Electric Power Research Institute, EPRI, are analyzed. In the remaining stages, general and specific guidelines for information displays, in particular, display pages, formats, elements and data quality and update rate recommendations are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMMLAB.

Integrated System Validation Usability Questionnaire: Computerized Procedures

International Nuclear Information System (INIS)

Garcés, Ma. I.; Torralba, B.

2015-01-01

The Research and Development (R&D) project on “Theoretical and Methodological Approaches to Integrated System Validation of Control Rooms, 2014-2015”, in which the research activities described in this report are framed, has two main objectives: to develop the items for an usability methodology conceived as a part of the measurement framework for performance-based control room evaluation that the OECD Halden Reactor Project will test in the experiments planned for 2015; and the statistical analysis of the data generated in the experimental activities of the Halden Man-Machine Laboratory (HAMMLAB) facility, with previous usability questionnaires, in 2010 and 2011. In this report, the procedure designed to meet the first goal of the project is described, in particular, the process followed to identify the items related to operating procedures, both computer and paper-based, one of the elements to be included in the usability questionnaire. Three phases are performed, in the first one, the approaches developed by the United States Nuclear Regulatory Commission, NRC, are reviewed, the models used by the nuclear industry and their technical support organizations, mainly, the Electric Power Research Institute, EPRI, are analyzed, and scientist advances are also explored. In the remaining stages, general and specific guidelines for computerized and paper-based procedures are compared and criteria for the preliminary selection of the items that should be incorporated into the usability questionnaire are defined. This proposal will be reviewed and adapted by the Halden Reactor Project to the design of the specific experiments performed in HAMLAB.

New Predictive Model at 11+0 to 13+6 Gestational Weeks for Early-Onset Preeclampsia With Fetal Growth Restriction.

Science.gov (United States)

Chang, Ying; Chen, Xu; Cui, Hong-Yan; Li, Xing; Xu, Ya-Ling

2017-05-01

The aim of the present study was to determine a predictive model for early-onset preeclampsia with fetal growth restriction (FGR) to be used at 11 +0 to 13 +6 gestational weeks, by combining the maternal serum level of pregnancy-associated plasma protein-A (PAPP-A), placental growth factor (PLGF), placental protein 13 (PP13), soluble endoglin (sEng), mean arterial pressure (MAP), and uterine artery Doppler. This was a retrospective cohort study of 4453 pregnant women. Uterine artery Doppler examination was conducted in the first trimester. Maternal serum PAPP-A, PLGF, PP13, and sEng were measured. Mean arterial pressure was obtained. Women were classified as with/without early-onset preeclampsia, and women with preeclampsia were classified as with/without FGR. Receiver operating characteristic analysis was performed to determine the value of the model. There were 30 and 32 pregnant women with early-onset preeclampsia with and without FGR. The diagnosis rate of early-onset preeclampsia with FGR was 67.4% using the predictive model when the false positive rate was set at 5% and 73.2% when the false positive rate was 10%. The predictive model (MAP, uterine artery Doppler measurements, and serum biomarkers) had some predictive value for the early diagnosis (11 +0 to 13 +6 gestational weeks) of early-onset preeclampsia with FGR.

Correlation of VCAM-1 expression in serum, cord blood, and placental tissue with gestational hypertension associated with fetal growth restriction in women from Xingtai Hebei, China.

Science.gov (United States)

Zhang, H G; Guo, W; Gu, H F; Chen, S B; Wang, J Q; Qiao, Z X; Ma, H S; Geng, S X

2016-08-26

The aim of this study was to investigate the expression of vascular adhesion molecule (VCAM)-1 in the maternal serum, cord blood, and placental tissue of pregnant women from Xingtai, Hebei, with gestational hypertension (GH) combined with fetal growth restriction (FGR). A total of 108 patients with GH combined with FGR (GH-FGR), 60 patients with GH alone (GH), and 50 healthy pregnant women (control) were recruited to this study. VCAM- 1 expression was detected in the maternal serum and cord blood by enzyme-linked immunosorbent assay, and in the placental tissue by immunohistochemistry. VCAM-1 expression was significantly higher in the maternal serum of patients with GH-FGR (164.38 ± 60.35) and GH alone (103.85 ± 54.47) than in the serum of the control population (46.70 ± 21.79; P 0.05). Moreover, the VCAM-1 expression rates were significantly higher and lower in the vascular endothelial and trophoblastic cells of the placenta of patients with GH-FGR (74.71 and 56.1%) and GH (72.98 and 55.36%), respectively, compared to those in the control subjects (46.48 and 95.11%). Therefore, we concluded that VCAM- 1 plays an important role in the development and generation of GH. Additionally, the low VCAM-1 expression in the trophoblastic cell could be correlated to the pathogenesis and progression of GH.

An Investigation of an Irradiated Fuel Pin by Measurement of the Production of Fast Neutrons in a Thermal Column and by Pile Oscillation Technique

Energy Technology Data Exchange (ETDEWEB)

Gustavsson, Veine

1968-05-15

A fuel pin irradiated to about 3400 MWd/tU from the Halden reactor has been investigated by a measurement of the production of fast neutrons in a thermal column and by pile oscillator technique in the central channel of the reactor R1. Calibration was made by using samples with different U 235 enrichments. The thermal column experiment gives the quantity ave({nu}{sigma}{sub f}) (average in the thermal column spectrum) for the Halden sample. {sigma}{sub f} is the macroscopic fission cross section and {nu} is the number of fast neutrons produced per fission. The result of the oscillator measurements is a value of ave({sigma}{sub a}) - w ave({sigma}{sub f}) (average in the central channel spectrum) for the irradiated sample, w is the importance of a fast neutron relative to a thermal one and ave({sigma}{sub a}) is the macroscopic absorption cross section. The results from both the experiments have been compared with values calculated by the REBUS code and the agreement was good.

Improving safety through an integrated approach for advanced control room development

International Nuclear Information System (INIS)

Haugset, K.; Berg, O.; Bologna, S.; Foerdestroemmen, N.T.; Kvalem, J.; Nelson, W.R.; Yamane, N.

1992-01-01

With the fast development of computer technology, the potential exists for improving operational safety of nuclear plants by using advanced operator tools in the control room. Specific systems are being introduced, such as systems for alarm handling, failure detection, disturbance diagnosis, procedural advice and others, often based on process modeling techniques or expert system technology. To ensure a maximum benefit from the new technology, a careful integration of the various systems must, however, take place, resulting in a well coordinated interface between the operator and the process. The OECD Halden Reactor Project has started the development of an Integrated Surveillance And Control System (ISACS). The basis for the activity is the experience at Halden in developing specific Computerized Operator Support Systems (COSSs), and the activity around the experimental control room HAMMLAB where detailed validations of operator tools have been performed for a number of years. The first goal in the ISACS project is to have a first, limited prototype in operation at the end of 1990. Validation experiments will follow. (orig.)

Improving safety through an integrated approach for advanced control room development

International Nuclear Information System (INIS)

Haugset, K.; Berg, O.; Foerdestroemmen, N.T.; Kvalem, J.; Nelson, W.R.

1990-01-01

With the fast development of computer technology, the potential exists for improving operational safety of nuclear plants by using advanced operator tools in the control room. Specific systems are being introduced, such as systems for alarm handling, failure detection, disturbance diagnosis, procedural advice and others, often based on process modeling techniques or expert system technology. To ensure a maximum benefit from the new technology, a careful integration of the various systems must, however, take place, resulting in a well coordinated interface between the operator and the process. The OECD Halden Reactor Project has started the development of an Integrated Surveillance And Control System (ISACS). The basis for the activity is the experience at Halden in developing specific Computerized Operator Support Systems (COSSs), and the activity around the experimental control room HAMMLAB where detailed validations of operator tools have been performed for a number of years. The first goal in the ISACS project is to have a first, limited prototype in operation at the end of 1990. Validation experiments will follow

Integrated process status overview

International Nuclear Information System (INIS)

Gertman, D.I.; Gaudio, P. Jr.

1986-01-01

This report summarizes findings to date with the IPSO, a large plant status overview currently under development at the OECD Halden Reactor Project. As part of a joint Halden and Combustion Engineering project, the overview is being tested in part to determine whether the large screen overview concept being entertained for use in the nuclear power plant (NPP) industry will facilitate operator performance. To this end an interactive simulation technique was used to establish a proof-of-principle test for the IPSO. Process control, operations, and human factors experts at Halden participated in the test and evaluation

Placental determinants of fetal growth: identification of key factors in the insulin-like growth factor and cytokine systems using artificial neural networks

Directory of Open Access Journals (Sweden)

Faleschini Elena

2008-06-01

Full Text Available Abstract Background Changes and relationships of components of the cytokine and IGF systems have been shown in placenta and cord serum of fetal growth restricted (FGR compared with normal newborns (AGA. This study aimed to analyse a data set of clinical and biochemical data in FGR and AGA newborns to assess if a mathematical model existed and was capable of identifying these two different conditions in order to identify the variables which had a mathematically consistent biological relevance to fetal growth. Methods Whole villous tissue was collected at birth from FGR (N = 20 and AGA neonates (N = 28. Total RNA was extracted, reverse transcribed and then real-time quantitative (TaqMan RT-PCR was performed to quantify cDNA for IGF-I, IGF-II, IGFBP-1, IGFBP-2 and IL-6. The corresponding proteins with TNF-α in addition were assayed in placental lysates using specific kits. The data were analysed using Artificial Neural Networks (supervised networks, and principal component analysis and connectivity map. Results The IGF system and IL-6 allowed to predict FGR in approximately 92% of the cases and AGA in 85% of the cases with a low number of errors. IGF-II, IGFBP-2, and IL-6 content in the placental lysates were the most important factors connected with FGR. The condition of being FGR was connected mainly with the IGF-II placental content, and the latter with IL-6 and IGFBP-2 concentrations in placental lysates. Conclusion These results suggest that further research in humans should focus on these biochemical data. Furthermore, this study offered a critical revision of previous studies. The understanding of this system biology is relevant to the development of future therapeutical interventions possibly aiming at reducing IL-6 and IGFBP-2 concentrations preserving IGF bioactivity in both placenta and fetus.

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

Science.gov (United States)

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

2018-02-01

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

Reaction behavior of SO2 in the sintering process with flue gas recirculation.

Science.gov (United States)

Yu, Zhi-Yuan; Fan, Xiao-Hui; Gan, Min; Chen, Xu-Ling; Chen, Qiang; Huang, Yun-Song

2016-07-01

The primary goal of this paper is to reveal the reaction behavior of SO2 in the sinter zone, combustion zone, drying-preheating zone, and over-wet zone during flue gas recirculation (FGR) technique. The results showed that SO2 retention in the sinter zone was associated with free-CaO in the form of CaSO3/CaSO4, and the SO2 adsorption reached a maximum under 900ºC. SO2 in the flue gas came almost from the combustion zone. One reaction behavior was the oxidation of sulfur in the sintering mix when the temperature was between 800 and 1000ºC; the other behavior was the decomposition of sulfite/sulfate when the temperature was over 1000ºC. However, the SO2 adsorption in the sintering bed mainly occurred in the drying-preheating zone, adsorbed by CaCO3, Ca(OH)2, and CaO. When the SO2 adsorption reaction in the drying-preheating zone reached equilibrium, the excess SO2 gas continued to migrate to the over-wet zone and was then absorbed by Ca(OH)2 and H2O. The emission rising point of SO2 moved forward in combustion zone, and the concentration of SO2 emissions significantly increased in the case of flue gas recirculation (FGR) technique. Aiming for the reuse of the sensible heat and a reduction in exhaust gas emission, the FGR technique is proposed in the iron ore sintering process. When using the FGR technique, SO2 emission in exhaust gas gets changed. In practice, the application of the FGR technique in a sinter plant should be cooperative with the flue gas desulfurization (FGD) technique. Thus, it is necessary to study the influence of the FGR technique on SO2 emissions because it will directly influence the demand and design of the FGD system.

Dosimetric Significance of the ICRP's Updated Guidance and Models, 1989-2003, and Implications for U.S. Federal Guidance

Energy Technology Data Exchange (ETDEWEB)

Leggett, R.W.

2003-09-10

Over the past two decades the U.S. Environmental Protection Agency (EPA) has issued a series of Federal guidance documents for the purpose of providing the Federal and State agencies with technical information to assist their implementation of radiation protection programs. Currently recommended dose conversion factors, annual limits on intake, and derived air concentrations for intake of radionuclides are tabulated in Federal Guidance Report No. 11 (FGR 11), published in 1988. The tabulations in FGR 11 were based on dosimetric quantities and biokinetic and dosimetric models of the International Commission on Radiological Protection (ICRP) developed for application to occupational exposures. Since the publication of FGR 11 the ICRP has revised some of its dosimetric quantities and its models for workers and has also developed age-specific models and dose conversion factors for intake of radionuclides by members of the public. This report examines the extent of the changes in the inhalation and ingestion dose coefficients of FGR 11 implied by the updated recommendations of the ICRP, both for workers and members of the public.

Diagnostic accuracy of fundal height and handheld ultrasound-measured abdominal circumference to screen for fetal growth abnormalities

Science.gov (United States)

Haragan, Adriane F.; Hulsey, Thomas C.; Hawk, Angela F.; Newman, Roger B.; Chang, Eugene Y.

2015-01-01

OBJECTIVE We sought to compare fundal height and handheld ultrasound–measured fetal abdominal circumference (HHAC) for the prediction of fetal growth restriction (FGR) or large for gestational age. STUDY DESIGN This was a diagnostic accuracy study in nonanomalous singleton pregnancies between 24 and 40 weeks’ gestation. Patients underwent HHAC and fundal height measurement prior to formal growth ultrasound. FGR was defined as estimated fetal weight less than 10%, whereas large for gestational age was defined as estimated fetal weight greater than 90%. Sensitivity and specificity were calculated and compared using methods described elsewhere. RESULTS There were 251 patients included in this study. HHAC had superior sensitivity and specificity for the detection of FGR (sensitivity, 100% vs 42.86%) and (specificity, 92.62% vs 85.24%). HHAC had higher specificity but lower sensitivity when screening for LGA (specificity, 85.66% vs 66.39%) and (sensitivity, 57.14% vs 71.43%). CONCLUSION HHAC could prove to be a valuable screening tool in the detection of FGR. Further studies are needed in a larger population. PMID:25818672

Role of the placental Vitamin D receptor in modulating feto-placental growth in Fetal growth restriction and Preeclampsia-affected pregnancies.

Directory of Open Access Journals (Sweden)

Padma eMurthi

2016-02-01

Full Text Available Fetal growth restriction (FGR is a common pregnancy complication that affects up to 5% of pregnancies worldwide. Recent studies demonstrate that Vitamin D deficiency is implicated in reduced fetal growth, which may be rescued by supplementation of Vitamin D. Despite this, the pathway(s by which Vitamin D modulate fetal growth remains to be investigated. Our own studies demonstrate that the Vitamin D receptor (VDR is significantly decreased in placentae from human pregnancies complicated by FGR and contributes to abnormal placental trophoblast apoptosis and differentiation and regulation of cell-cycle genes in vitro. Thus, Vitamin D signalling is important for normal placental function and fetal growth. This review discusses the association of Vitamin D with fetal growth, the function of Vitamin D and its receptor in pregnancy, as well as the functional significance of a placental source of Vitamin D in FGR. Additionally, we propose that for Vitamin D to be clinically effective to prevent and manage FGR, the molecular mechanisms of Vitamin D and its receptor in modulating fetal growth requires further investigation.

Prediction of Fetal Growth Restriction by Analyzing the Messenger RNAs of Angiogenic Factor in the Plasma of Pregnant Women.

Science.gov (United States)

Takenaka, Shin; Ventura, Walter; Sterrantino, Anna Freni; Kawashima, Akihiro; Koide, Keiko; Hori, Kyoko; Farina, Antonio; Sekizawa, Akihiko

2015-06-01

To predict the occurrence of fetal growth restriction (FGR) by analyzing messenger RNA (mRNA) expression levels of vascular endothelial growth factor receptor 1 (fms-like tyrosine kinase 1 [Flt-1]) in maternal blood. Eleven women with FGR were matched with 88 controls. Plasma samples were obtained during each trimester. The Flt-1 mRNA expression levels were compared between groups. Predicted probabilities were calculated, and sensitivity-specificity (receiver-operating characteristic [ROC]) curves were assessed based on regression models for each trimester measurement and possible combinations of measurements. The mRNA levels of the FGR group during all trimesters were significantly higher than those of the control group. The ROC curve of combined first and second trimester data yielded a detection rate of 60% at a 10% false-positive rate, with an area under curve of 0.79. The Flt-1 mRNA expression in maternal blood can be used as a marker to predict the development of FGR, long before a clinical diagnosis is made. © The Author(s) 2014.

The task complexity experiment 2003/2004

International Nuclear Information System (INIS)

Laumann, Karin; Braarud, Per Oeivind; Svengren, Haakan

2005-08-01

The purpose of this experiment was to explore how additional tasks added to base case scenarios affected the operators' performance of the main tasks. These additional tasks were in different scenario variants intended to cause high time pressure, high information load, and high masking. The experiment was run in Halden Man-Machine Laboratory's BWR simulator. Seven crews participated, each for one week. There were three operators in each crew. Five main types of scenarios and 20 scenario variants were run. The data from the experiment were analysed by completion time for important actions and by in-depth qualitative analyses of the crews' communications. The results showed that high time pressure decreased some of the crews' performance in the scenarios. When a crew had problems in solving a task for which the time pressure was high, they had even more problems in solving other important tasks. High information load did not affect the operators' performance much and in general the crews were very good at selecting the most important tasks in the scenarios. The scenarios that included both high time pressure and high information load resulted in more reduced performance for the crews compared to the scenarios that only included high time pressure. The total amount of tasks to do and information load to attend to seemed to affect the crews' performance. To solve the scenarios with high time pressure well, it was important to have good communication and good allocation of tasks within the crew. Furthermore, the results showed that scenarios with an added complex, masked task created problems for some crews when solving a relatively simple main task. Overall, the results confirmed that complicating, but secondary tasks, that are not normally taken into account when modelling the primary tasks in a PRA scenario can adversely affect the performance of the main tasks modelled in the PRA scenario. (Author)

A new reactor core monitoring system. First experience gained at the Dukovany NPP

International Nuclear Information System (INIS)

Pecka, M.; Svarny, J.; Kment, J.

2001-01-01

The article deals with methods of interpretation of in-core measurements that are based on the determination of the three-dimensional (3D) power distribution within the reactor core, discusses on-line mode calculations, and describes the results obtained during the trial operation of the new SCORPIO-VVER reactor core monitoring system. The principles of the method of determination of the fuel assembly subchannel parameters are outlined. Alternative methods of self-powered detector signal conversion to local power are given, and some results of their testing are presented. Emphasis is put on self-powered detectors supplied by the US firm IST, which were first deployed at the Dukovany NPP in 1998. The predictive function of the SCORPIO-VVER system, whose implementation was inspired by favourable experience gained on some PWR reactors (such as the products of the Halden reactor project at Ringhals and Sizewell B) were adapted to the specific needs of WWER-440 reactors. The main results of validation of the functions are described and presented in detail. (author)

Differential correlations between maternal hair levels of tobacco and alcohol with fetal growth restriction clinical subtypes.

Science.gov (United States)

Sabra, Sally; Malmqvist, Ebba; Almeida, Laura; Gratacos, Eduard; Gomez Roig, Maria Dolores

2018-08-01

Maternal exposure to tobacco and alcohol is a known cause, among others, for fetal growth restriction (FGR). Clinically, FGR can be subclassified into two forms: intrauterine growth restriction (IUGR) and small for gestational age (SGA), based on the severity of the growth retardation, and abnormal uterine artery Doppler or cerebro-placental ratio. This study aimed at investigating any differential correlation between maternal exposures to these toxins with the two clinical forms of FGR. Therefore, a case-control study was conducted in Barcelona, Spain. Sixty-four FGR subjects, who were further subclassified into IUGR (n = 36) and SGA (n = 28), and 89 subjects matched appropriate-for-gestational age (AGA), were included. The levels of nicotine (NIC) and ethyl glucuronide (EtG), biomarkers of tobacco and alcohol exposure, respectively, were assessed in the maternal hair in the third trimester. Our analysis showed 65% of the pregnant women consumed alcohol, 25% smoked, and 19% did both. The odds ratios (ORs) of IUGR were 21 times versus 14 times for being SGA with maternal heavy smoking, while with alcohol consumption the ORs for IUGR were 22 times versus 37 times for the SGA group. The differential correlations between these toxins with the two subtypes of FGR suggest different mechanisms influencing fetal weight. Our alarming data of alcohol consumption during pregnancy should be considered for further confirmation among Spanish women. Copyright © 2018 Elsevier Inc. All rights reserved.

Experiences from the development of the FAME communication manager using the CASE-tool SDT

International Nuclear Information System (INIS)

Stoelen, Ketil; Mohn, Peter; Sandmark, Haakon; Thunem, Harald

1999-05-01

The three year programme 1997-1999 for the OECD Halden Reactor Project (HRP) identifies the need to gain experience from applying formal techniques in real-life system developments. This motivated the initiation of the HRP research activity Integration of Formal Specification in the Development of HAMMLAB 2000 (INT-FS). The principal objective was to experiment with formal techniques in system developments at the HRP; in particular, system developments connected to HAMMLAB 2000 - the computerised laboratory for man-machine-interaction experiments currently under construction. It was hoped that this experimentation with formal techniques should result in a better understanding of how such techniques should be utilised in a more industrial setting. To obtain more knowledge with respect to the practical effects and consequences of an increased level of formalization was another objective. This report presents the results from an INT-FS development of a control component. The report describes the architecture of this component, the techniques, methods and tools used during its development, and the background of the personnel taking part. It also outlines the development process and describes the activities within the various development stages. Finally, it summarizes experiences and results, and discusses their relevance. The report concludes that the selected formalisms and tools are helpful for the development of distributed systems. In particular, the formalization of the informal requirements identified many weaknesses and inconsistencies. The involved parties found the description techniques easy to understand and well-suited as a medium for discussing and capturing requirements. The simulation of design was also a very positive experience (author) (ml)

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Science.gov (United States)

González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

2015-10-01

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

Amnioinfusion before 26 weeks' gestation for severe fetal growth restriction with oligohydramnios: preliminary pilot study.

Science.gov (United States)

Takahashi, Yuichiro; Iwagaki, Shigenori; Chiaki, Rika; Iwasa, Tomotake; Takenaka, Motoki; Kawabata, Ichiro; Itoh, Mitsuaki

2014-03-01

The prognosis for severe fetal growth restriction (FGR) with severe oligohydramnios before 26 weeks' gestation (WG) is currently poor; furthermore, its management is controversial. We report the innovative new management of FGR, such as therapeutic amnioinfusion and tocolysis. For FGR and severe oligohydramnios before 26 WG complicated with absent or reversed umbilical artery end-diastolic flow velocity and/or deceleration by ultrasonography, we performed transabdominal amnioinfusion with tocolysis. Cases with multiple anomalies were excluded. Survival rate and long-term prognosis were analyzed. Among 570 FGR cases, 18 were included in the study. Mean diagnosis and delivery were at 22.6 ± 2.0 and 28.7 ± 3.3 WG. Median birthweight was 625 g (-4.2 standard deviation). Final survival rate was 11/13 (85%). There were five fetal deaths. In seven cases, oligohydramnios improved. Growth was detected in 10/18 fetuses. Furthermore, 8/8 decelerations, 4/12 cases of reversed umbilical artery end-diastolic flow velocity, 7/14 cases of brain-sparing effect, and 6/13 venous Doppler abnormalities were improved. When we detected umbilical cord compression, 8/10 cases were rescued. Eleven infants were followed up for an average of 5 years; one case of cerebral palsy with normal development and 10 cases with intact motor functions without major neurological handicap were confirmed. In cases of extremely severe FGR before 26 WG with oligohydramnios and circulatory failure, amnioinfusion might be a promising, innovative tool. © 2013 The Authors. Journal of Obstetrics and Gynaecology Research © 2013 Japan Society of Obstetrics and Gynecology.

Circulating cell-derived microparticles in severe preeclampsia and in fetal growth restriction.

Science.gov (United States)

Alijotas-Reig, Jaume; Palacio-Garcia, Carles; Farran-Codina, Immaculada; Ruiz-Romance, Mar; Llurba, Elisa; Vilardell-Tarres, Miquel

2012-02-01

The behavior of the circulating microparticles (cMP) in severe preeclampsia (PE) and fetal growth restriction (FGR) is disputed. METHOD OF STUDY  Non-matched case-control study. Seventy cases of severe PE/HELLP/FGR were compared to 38 healthy pregnant women. Twenty healthy non-pregnant women acted as a control. cMP were analyzed using flow cytometry. Results are given as total (annexin-A5-ANXA5+), platelet (CD41+), leukocyte (CD45+), endothelial (CD144+CD31+//CD41-), and CD41-negative cMP/μL of plasma. Antiphospholipid antibodies (aPL) were analyzed through usual methods. Platelet and endothelial cMP increased in healthy pregnant women. PE whole group (PE±FGR) showed an increase in endothelial and CD41-negative, but not in platelet-derived, cMP. Comparing PE whole group versus healthy pregnant, we found cMP levels of endothelial and CD41- had increased. The cMP results obtained in PE group were similar to those of the PE whole group. Comparing PE group to isolated FGR, significant CD41-negative cMP increase was found in PE. According to its aPL positivity, a trend to decrease in leukocyte and endothelial-derived cMP was found in PE group. Normal pregnancy is accompanied by endothelial and platelet cell activation. Endothelial cell activation has been shown in PE but not in isolated FGR. In PE, aPL may contribute to endothelial and possibly to leukocyte cell activation. © 2011 John Wiley & Sons A/S.

HOTLAB: European hot laboratories research capacities and needs. Plenary meeting 2004

International Nuclear Information System (INIS)

Oberlaender, B.C.; Jenssen, H.K.

2005-01-01

The report presents proceedings from the 2004 annual HOTLAB plenary meeting at Halden and Kjeller, Norway. The goal of the yearly plenary meeting was to: Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research. Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling, etc. Promote normalisation and co-operation, e.g. by looking at mutual complementarities. Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The main themes of the five topical oral sessions of the Halden plenary meeting cover: Work package leaders report and specific papers, presentation of PIE facility databases, i.e. one worldwide (IAEA) and one inside the European communities. Reports from present and future needs and on nuclear transports. Refabrication and instrumentation: Available equipment, technical characteristics such as fabrication procedures, hot-cell compatibility, and practical experiences. Post irradiation examination: Updated and new remote techniques and methodologies, new materials such as inert matrix fuels, spallation sources and neutron absorber materials. Refurbishment and decommissioning: reports on refurbishment and decommissioning of PIE facilities. Waste and transport: Hot laboratory waste characteristics and handling, spent fuel research. Several posters are presented

HOTLAB: European hot laboratories research and capacities and needs. Plenary meeting 2004

Energy Technology Data Exchange (ETDEWEB)

Oberlaender, B.C.; Jenssen, H.K. (ed.)

2005-01-01

The report presents proceedings from the 2004 annual HOTLAB plenary meeting at Halden and Kjeller, Norway. The goal of the yearly plenary meeting was to: Exchange experience on analytical methods, their implementation in hot cells, the methodologies used and their application in nuclear research. Share experience on common infrastructure exploitation matters such as remote handling techniques, safety features, QA-certification, waste handling, etc. Promote normalisation and co-operation, e.g. by looking at mutual complementarities. Prospect present and future demands from the nuclear industry and to draw strategic conclusions regarding further needs. The main themes of the five topical oral sessions of the Halden plenary meeting cover: Work package leaders report and specific papers, presentation of PIE facility databases, i.e. one worldwide (IAEA) and one inside the European communities. Reports from present and future needs and on nuclear transports. Refabrication and instrumentation: Available equipment, technical characteristics such as fabrication procedures, hot-cell compatibility, and practical experiences. Post irradiation examination: Updated and new remote techniques and methodologies, new materials such as inert matrix fuels, spallation sources and neutron absorber materials. Refurbishment and decommissioning: reports on refurbishment and decommissioning of PIE facilities. Waste and transport: Hot laboratory waste characteristics and handling, spent fuel research. Several posters are presented.

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

Science.gov (United States)

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

2017-11-01

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

Views on quality assurance at Finnish and Swedish nuclear power plants and at Halden Reactor; Syn paa kvalitetssaekring vid finlaendska och svenska kaernkraftverk samt vid Haldenreaktorn

Energy Technology Data Exchange (ETDEWEB)

Hammar, L.; Lidh, B. [ES-konsult (Sweden); Wahlstroem, B.; Reiman, T. [VTT Automation (Finland)

2001-06-01

The paper reports on a study within the Nordic Nuclear Safety Research, NKS on quality systems at nuclear installations in Finland, Norway and Sweden. In the study a total of 74 people at the NPPs in Barsebaeck, Forsmark, Loviisa, Olkiluoto, Oskarshamn and Ringhals, and at the research reactor in Halden were interviewed in the period 30 August to 13 December 2000 concerning their views in regard of quality and quality systems. The study was concluded with a seminar held in the Ringhals nuclear power plant in Januar 2001. The study covered a number of aspects in regard of quality management, including the quality concept, quality systems, topical quality issues and approaches, rules and procedures, competency and training, the process approach to quality management, the promotion of quality consciousness and future prospects. The study reflects the significant progress made in the management of quality in nuclear power in the Nordic countries since the early phase in the seventies. The most distinctive characteristic of today's approach to quality is seen in that responsibility for the quality is assumed directly in conjunction with the working processes. It could be noted that the work patterns at the nuclear installations have been largely modified during the recent years as a result of persistent endeavours to continuously improve the quality of operation. Challenges were seen in currently reduced revenues due to descending electricity prices and the likely prospect of further increased regulatory safety requirements. The report is aimed for those working with quality issues at the nuclear power plants as well as for those interested in quality management in general or in the safety aspects of nuclear power in particular. (au)

Modelling of LOCA Tests with the BISON Fuel Performance Code

Energy Technology Data Exchange (ETDEWEB)

Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

2016-05-01

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

Recommendations to alarm systems and lessons learned on alarm system implementation

International Nuclear Information System (INIS)

Soerenssen, Aimar; Veland, Oeystein; Farbrot, Jan Erik; Kaarstad, Magnhild; Seim, Lars Aage; Foerdestroemmen, Nils; Bye, Andreas

2001-11-01

Alarm systems have been of major concern within complex industrial processes for many years. Within the nuclear community, the TMI accident in 1979 was the first really serious event that showed also the importance of the man-machine aspects of the systems in general, and the alarm system in particular. The OECD Halden Reactor Project has been working with alarm systems since 1974. This report is an attempt to gather some of the knowledge that has been accumulated during the years in Halden, both in research and also in bilateral projects. Bilateral projects within this field have provided a practical basis of knowledge.A major part of this report consists of a set of recommendations, which reflect HRP's current understanding of how an alarm system should work. There are also recommendations on design methods. But also other issues are included, as system development and implementation experience, and experimental knowledge on the performance of alarm systems. Some open issues are also discussed. (Author). 54 refs., 15 figs

Continuous improvement of software quality

International Nuclear Information System (INIS)

Sivertsen, Terje

1999-04-01

The present report is the first Halden Work Report delivered from the OECD Halden Reactor Project's research activity on formal methods and software quality. Of particular concern in this activity is to reach a consensus between regulators, licensees and the nuclear industry on questions related to the effective, industrial use of formal methods. The report gives considerable attention to the importance of continuous improvement as a characteristic of a living software quality system, and to the need of providing a basis for software process/product quality integration. In particular, the report discusses these aspects from the perspectives of defect prevention, formal methods, Total Quality Management (TQM), and Bayesian Belief Nets. Another concern is to promote controlled experiments on the use of new methods, techniques, and tools. This is achieved partly by reviewing suggestions on the collection and experimental use of data, and by surveying a number of metrics believed to have some potential for comparison studies (author) (ml)

A Study of the Temperature Distribution in UO{sub 2} Reactor Fuel Elements

Energy Technology Data Exchange (ETDEWEB)

Devold, I

1968-05-15

Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO{sub 2} fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP.

A Study of the Temperature Distribution in UO2 Reactor Fuel Elements

International Nuclear Information System (INIS)

Devold, I.

1968-05-01

Thermal conductivity is one of the most important properties of nuclear reactor fuels. Accurate knowledge of this property is vital because, among other things, it determines the maximum power that can be taken out of the fuel element per unit length of the material without exceeding the safety limits of the fuel elements. This report consists of a study of the thermal behaviour of uranium dioxide in the form of reactor fuel. The experimental part of the report describes measurements performed at the OECD Halden Reactor Project, Halden, Norway. The experiment was originally set up in order to measure the temperature at the center of a UO 2 fuel element as a function of element power, in order to determine the safe operation limit of the fuel assembly. However, in analysing the data obtained, very interesting thermal conductivity values were obtained and comparison with existing correlations could be performed. This comparison shows that a certain agreement is obtained between the measured data at Halden and a theory published by J.L. Bates in 1961, which predicts an increase in the thermal conductivity above 1500 deg C. The data obtained below 1300 deg C are also in good agreement with measurements performed by Vogt, Grandell and Runfors in 1964. The report contains a mathematical description of the heat transfer mechanisms in cylindrical fuel elements. The model is coded in FORTRAN IV-code and referred to as FUELTEMP

Irradiation performance and post-irradiation examinations of the instrumented sphere-pac UO2 assembly IFA-204, irradiated up to 1.7% FIMA in the Halden Boiling Water Reactor

International Nuclear Information System (INIS)

Linde, A. van der.

1982-12-01

Fuel assembly IFA-204 consisted of four pins filled with a blend of three sizes of UO 2 spheres. This blend was compacted in the zircaloy-4 cladding tube by vibration to achieve a 87-88% T.D. smear density of the 1500 mm long sphere-pac fuel column. IFA-204 was irradiated during 597 days, equivalent to 436 full power days, in the Halden BWR from November 1971 to April 1974 when the achieved assembly average burnup amounted to 13.7 MWd/kg UO 2 . Two of the four pins were equipped with fuel column length and pin length extensometers. The measurements showed that at average pin powers less than about 25 kW.m -1 only fuel-clad-thermal-interaction, FCTI, occurred. The clad thermal expansion coefficient was 4.8 ppm/kW.m -1 . For calculations of the irradiation behaviour of sphere-pac LWR pins, filled with 0.1 MPa helium, with the Gapcon-Thermal-2 code a new thermal conductivity-temperature relationship has been developed. The calculated data agreed reasonably well with the measured data when taking into account a restructured fuel density of 90% T.D., a fuel surface roughness of 1 μm, the absence of a fuel-cladding gap and an effective full power fuel restructuring time of 40 days. It is concluded that the porous structure of the sphere-pac fuel column, whether restructured or not, has the inherent disadvantage of a high mobility of gaseous fission products and the inherent advantage of a practically stress free operation of the cladding

Procedure automation: the effect of automated procedure execution on situation awareness and human performance

International Nuclear Information System (INIS)

Andresen, Gisle; Svengren, Haakan; Heimdal, Jan O.; Nilsen, Svein; Hulsund, John-Einar; Bisio, Rossella; Debroise, Xavier

2004-04-01

As advised by the procedure workshop convened in Halden in 2000, the Halden Project conducted an experiment on the effect of automation of Computerised Procedure Systems (CPS) on situation awareness and human performance. The expected outcome of the study was to provide input for guidance on CPS design, and to support the Halden Project's ongoing research on human reliability analysis. The experiment was performed in HAMMLAB using the HAMBO BWR simulator and the COPMA-III CPS. Eight crews of operators from Forsmark 3 and Oskarshamn 3 participated. Three research questions were investigated: 1) Does procedure automation create Out-Of-The-Loop (OOTL) performance problems? 2) Does procedure automation affect situation awareness? 3) Does procedure automation affect crew performance? The independent variable, 'procedure configuration', had four levels: paper procedures, manual CPS, automation with breaks, and full automation. The results showed that the operators experienced OOTL problems in full automation, but that situation awareness and crew performance (response time) were not affected. One possible explanation for this is that the operators monitored the automated procedure execution conscientiously, something which may have prevented the OOTL problems from having negative effects on situation awareness and crew performance. In a debriefing session, the operators clearly expressed their dislike for the full automation condition, but that automation with breaks could be suitable for some tasks. The main reason why the operators did not like the full automation was that they did not feel being in control. A qualitative analysis addressing factors contributing to response time delays revealed that OOTL problems did not seem to cause delays, but that some delays could be explained by the operators having problems with the freeze function of the CPS. Also other factors such as teamwork and operator tendencies were of importance. Several design implications were drawn

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

International Nuclear Information System (INIS)

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

2012-12-01

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

Elsevier Trophoblast Research Award Lecture: Searching for an early pregnancy 3-D morphometric ultrasound marker to predict fetal growth restriction.

Science.gov (United States)

Collins, S L; Stevenson, G N; Noble, J A; Impey, L

2013-03-01

Fetal growth restriction (FGR) is a major cause of perinatal morbidity and mortality, even in term babies. An effective screening test to identify pregnancies at risk of FGR, leading to increased antenatal surveillance with timely delivery, could decrease perinatal mortality and morbidity. Placental volume, measured with commercially available packages and a novel, semi-automated technique, has been shown to predict small for gestational age babies. Placental morphology measured in 2-D in the second trimester and ex-vivo post delivery, correlates with FGR. This has also been investigated using 2-D estimates of diameter and site of cord insertion obtained using the Virtual Organ Computer-aided AnaLysis (VOCAL) software. Data is presented describing a pilot study of a novel 3-D method for defining compactness of placental shape. We prospectively recruited women with a singleton pregnancy and BMI of Elsevier Ltd. All rights reserved.

EVERREST prospective study: a 6-year prospective study to define the clinical and biological characteristics of pregnancies affected by severe early onset fetal growth restriction.

Science.gov (United States)

Spencer, Rebecca; Ambler, Gareth; Brodszki, Jana; Diemert, Anke; Figueras, Francesc; Gratacós, Eduard; Hansson, Stefan R; Hecher, Kurt; Huertas-Ceballos, Angela; Marlow, Neil; Marsál, Karel; Morsing, Eva; Peebles, Donald; Rossi, Carlo; Sebire, Neil J; Timms, John F; David, Anna L

2017-01-23

Fetal growth restriction (FGR) is a serious obstetric condition for which there is currently no treatment. The EVERREST Prospective Study has been designed to characterise the natural history of pregnancies affected by severe early onset FGR and establish a well phenotyped bio-bank. The findings will provide up-to-date information for clinicians and patients and inform the design and conduct of the EVERREST Clinical Trial: a phase I/IIa trial to assess the safety and efficacy of maternal vascular endothelial growth factor (VEGF) gene therapy in severe early onset FGR. Data and samples from the EVERREST Prospective Study will be used to identify ultrasound and/or biochemical markers of prognosis in pregnancies with an estimated fetal weight (EFW) economic impact; psychological impact; neonatal condition, progress and complications; and infant growth and neurodevelopment to 2 years of corrected age in surviving infants. Standardised longitudinal ultrasound measurements are performed, including: fetal biometry; uterine artery, umbilical artery, middle cerebral artery, and ductus venosus Doppler velocimetry; and uterine artery and umbilical vein volume blood flow. Samples of maternal blood and urine, amniotic fluid (if amniocentesis performed), placenta, umbilical cord blood, and placental bed (if caesarean delivery performed) are collected for bio-banking. An initial analysis of maternal blood samples at enrolment is planned to identify biochemical markers that are predictors for fetal or neonatal death. The findings of the EVERREST Prospective Study will support the development of a novel therapy for severe early onset FGR by describing in detail the natural history of the disease and by identifying women whose pregnancies have the poorest outcomes, in whom a therapy might be most advantageous. The findings will also enable better counselling of couples with affected pregnancies, and provide a valuable resource for future research into the causes of FGR. NCT02097667

Development and testing of an Internet-based data collection technique for simulator and real world experimentation

International Nuclear Information System (INIS)

Droeivoldsmo, Asgeir; Johnsen, Terje

2005-09-01

With experience from many years of data collection in the Man - Machine and Virtual Reality Laboratories at the OECD Halden Reactor Project, an evident need for more efficient handling of questionnaire data was documented. A working prototype on-line system for World Wide Web (www) questionnaire generation and data collection was developed and tested. This paper discusses the use of www-based data collection and the need for system functionality in experiments and surveys. Insights from the development of the system are reported together with experiences using such tools in simulation and realistic field experimentation. (Author)

Irradiation of UO2+x fuels in the TANOX device

International Nuclear Information System (INIS)

Dehaudt, P.; Caillot, L.; Delette, G.; Eminet, G.; Mocellin, A.

1998-01-01

The TANOX analytical irradiation device is presented and the first results concerning stoichiometric and hyper stoichiometric uranium dioxide fuels with two different grain sizes are given. The TANOX device is designed to obtain rapidly significant burnups in fuels at relatively low temperatures. It is placed at the periphery of the SILOE reactor and translated to adjust the irradiation power. The continuous measure of the centre-line temperature allows to control the experiment and to evaluate the thermal behaviour of the rods. A TANOX fuel rod has a length of 100 mm with 20 fuel pellets in a stainless steel cladding and is inserted in a thick aluminium alloy overcladding which is cooled by the primary water circuit reactor. These conditions of small size pellets and improved thermal exchanges have been designed to dissipate the heat power due to fission densities three to five times higher than in a PWR. The first analytical irradiation was devoted to the study of UO 2.00 , UO 2.01 and UO 2.02 fuels with standard and large grain sizes obtained by annealing. A burnup of about 9000 MWd.t -1 U was reached in these fuels. The thermal analysis shows a degraded conductivity for the UO 2.02 fuel rod due to the hyper stoichiometry. The released fractions of 85 Kr during irradiation are negligible as expected (lower than 0,1%). Some of the pellets were heat treated at 1700 deg. C for 5 hours. The gas release was analysed after 30 minutes and at the end of the treatment. The main results are as follows: the fission gas release (FGR) of the standard UO 2 varies from one sample to another; the FGR of the hyper stoichiometric fuels is of the same order of magnitude than that of the stoichiometric UO 2 fuel of normal grain sizes; the grain size increase has no effect on FGR for UO 2.00 but considerably decreases the FGR for UO 2.01 and UO 2.02 fuels. These heat treated samples are also observed to characterize the inter- and intragranular fission gas bubbles. (author)

Role of uteroplacental and fetal Doppler in identifying fetal growth restriction at term.

OpenAIRE

Khalil, A; Thilaganathan, B

2017-01-01

Identification of the fetus at risk of adverse outcome at term is a challenge to both clinicians and researchers alike. Despite the fact that fetal growth restriction (FGR) is a known risk factor for stillbirth, at least two thirds of the stillbirth cases at term are not small for gestational age (SGA) - a commonly used proxy for FGR. However, the majority of SGA fetuses are constitutionally small babies and do not suffer from adverse perinatal outcome. The cerebroplacental ratio (CPR) is eme...

Concept of the LORELEI Test Device for LOCA Experiment in the JHR Reactor

International Nuclear Information System (INIS)

Moran, N.; Ferry, L.; Azulay, A.; Mileguir, O.; Weiss, Y.; Szanto, M.

2014-01-01

Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis. Former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. In JHR material testing reactor, which is currently under construction, one significant experimental device is the LORELEI testing device. The objective is to examine the LOCA sequence influence on: thermo-mechanical behavior of the fuel clad, possible fuel relocation, corrosion at high temperature, oxidation, hydriding and resulted clad embrittment. The device is a single rod closed loop system placed on a displacement device inside a defined channel in the reflector. Several operational constrains on the device, as required by the reactor operational philosophy resulted quite a few challenges in the design. Constrains as: pre experimental re-irradiation phase under thermo-syphonic flow, application of active insulation to simulate the surrounding fuel, application of tensile force during refolding simulation, controlling the experiment with non-direct temperature measurement, etc. requires sophisticated solutions. The main objective of the conceptual design was to remove the uncertainties of those challenging requirements. The current presentation describes the approach applied defining the concept of the device, using sophisticated design combined with computational and experimental tools

Nuclear enterprises at the Institute for Energy Technology - IFE. A socio-economic cost/benefit analysis; Nukleaere virksomheter ved Institutt for energiteknikk - IFE. En samfunnsoekonomisk kost/nytte-analyse

Energy Technology Data Exchange (ETDEWEB)

2008-03-15

A cost-benefit analysis concerning the research reactors JEEP II at Kjeller and the Halden Reactor in Halden, operated by the Institute for Energy Technology. It is concluded for both of the reactors that the benefits of continued operations are outweigh the cost. Financing, accident risk, waste management and nuclear competence are some of the aspects treated. The Norwegian Ministry of Trade and Industry initiated the evaluation on behalf of the Norwegian Government

Experiences from the formal specification of the integration platform and the synthesis of SDT with the software bus

International Nuclear Information System (INIS)

Thunem, Harald; Mohn, Peter; Sandmark, Haakon; Stoelen, Ketil

1999-04-01

The three year programme 1997-1999 for the OECD Halden Reactor Project (HRP) identifies the need to gain experience from applying formal techniques in real-life system developments. This motivated the initiation of the HRP research activity Integration of Formal Specification in the Development of HAMMLAB 2000 (INT-FS). The principal objective was to experiment with formal techniques in system developments at the HRP; in particular, system developments connected to HAMMLAB 2000 - the computerised laboratory for man-machine-interaction experiments currently under construction. It was hoped that this experimentation with formal techniques should result in a better understanding of how such techniques should be utilised in a more industrial setting. To obtain more knowledge with respect to the practical effects and consequences of an increased level of formalization was another objective. This report summarises experiences, results and conclusions from a pre-study addressing INT-FS related issues connected to the development of the HAMMLAB 2000 Integration Platform (IP). The report starts by giving a brief overview of the IP. Then it describes and summarises experiences from the formalization of a top-level requirements specification for the IP. Finally, it discusses various approaches for the integration of applications generated automatically through the CASE-tool SDT and the Software Bus on which the communication within HAMMLAB 2000 will be based. The report concludes that the selected formalisms and tools are well-suited to describe IP-like systems. The report also concludes that the integration of SDT applications with the Software Bus will not be a major obstacle, and finally that a monitoring component for the IP is well-suited for development within INT-FS (author) (ml)

Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

Energy Technology Data Exchange (ETDEWEB)

González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Serrano-Purroy, D. [European Commission - EC, Joint Research Centre (JRC), Institute for Transuranium Elements - ITU, Postfach 2340, D-76125 Karlsruhe (Germany); Sureda, R. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Casas, I. [Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain); Pablo, J. de [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain)

2015-10-15

The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO{sub 2} spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAP{sub c}) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

OECD - HRP Summer School on Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

International Summer School on Nuclear Fuel

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

OECD - HRP Summer School on Nuclear Fuel

International Nuclear Information System (INIS)

2000-01-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures

Poorer detection rates of severe fetal growth restriction in women of likely refugee background: A case for re-focusing pregnancy care.

Science.gov (United States)

Biro, Mary Anne; East, Christine

2017-04-01

Severe fetal growth restriction (FGR) (Women of refugee background have been found to have poorer perinatal outcomes compared to others and these outcomes can in part be explained by previous history. However, less access to and engagement with pregnancy care may also be contributing factors. This study examined the impact of likely refugee background on severe FGR in a singleton pregnancy undelivered by 40 weeks. A retrospective study was undertaken utilising data on women who gave birth to a severely growth-restricted infant at Monash Health during January 2013-July 2015. Unadjusted and adjusted analyses were undertaken to examine the association between the mother being of likely refugee background and severe FGR in singletons delivered after 40 weeks. There was an association between the mother being of likely refugee background and giving birth to a severely growth-restricted baby after 40 weeks with these mothers at two and half times the odds compared to mothers of non-refugee background (adjusted odds ratio 2.52; 95% confidence interval: 1.44-4.42). While detecting FGR is clinically challenging, our findings suggest that maternity services need to be supported to offer care tailored to the specific needs of vulnerable and disadvantaged populations. Providing quality, culturally responsive and accessible care is fundamental to addressing refugee maternal and perinatal health inequalities. © 2017 The Royal Australian and New Zealand College of Obstetricians and Gynaecologists.

Down regulation of macrophage IFNGR1 exacerbates systemic L. monocytogenes infection.

Directory of Open Access Journals (Sweden)

Emily M Eshleman

2017-05-01

Full Text Available Interferons (IFNs target macrophages to regulate inflammation and resistance to microbial infections. The type II IFN (IFNγ acts on a cell surface receptor (IFNGR to promote gene expression that enhance macrophage inflammatory and anti-microbial activity. Type I IFNs can dampen macrophage responsiveness to IFNγ and are associated with increased susceptibility to numerous bacterial infections. The precise mechanisms responsible for these effects remain unclear. Type I IFNs silence macrophage ifngr1 transcription and thus reduce cell surface expression of IFNGR1. To test how these events might impact macrophage activation and host resistance during bacterial infection, we developed transgenic mice that express a functional FLAG-tagged IFNGR1 (fGR1 driven by a macrophage-specific promoter. Macrophages from fGR1 mice expressed physiologic levels of cell surface IFNGR1 at steady state and responded equivalently to WT C57Bl/6 macrophages when treated with IFNγ alone. However, fGR1 macrophages retained cell surface IFNGR1 and showed enhanced responsiveness to IFNγ in the presence of type I IFNs. When fGR1 mice were infected with the bacterium Listeria monocytogenes their resistance was significantly increased, despite normal type I and II IFN production. Enhanced resistance was dependent on IFNγ and associated with increased macrophage activation and antimicrobial function. These results argue that down regulation of myeloid cell IFNGR1 is an important mechanism by which type I IFNs suppress inflammatory and anti-bacterial functions of macrophages.

Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

2010-02-16

Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

Energy Technology Data Exchange (ETDEWEB)

Clark, Scarlett R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The

Early detection and diagnosis of disturbances in nuclear power plants

International Nuclear Information System (INIS)

Bjorlo, T.J.; Berg, O.; Grini, R.E.; Yokobayashi, M.

1987-01-01

The surveillance and control of nuclear power plants comprises a number of tasks and functions which have to be shared between the operators and the control and instrumentation systems. The trend in control room design towards a higher degree of computerization of the control and instrumentation systems and replacement of conventional instrument panels by VDU-based man-machine communication systems opens possibilities for improving the support given to the operators in their cognitive tasks. At the OECD Halden Reactor Project these possibilities are explored through a research and development programme centered around the NORS/HAMMLAB experimental control room facility. The full-scale PWR simulator, NORS, coupled with the HAlden Man-Machine LABoratory (HAMMLAB), which includes the experimental control room as well as an established research methodology and staff, constitutes a unique basis ofr the design, development and validation of operator support systems, as well as for more basic operator performance experimentation. The aim of the system development work at the Halden Project is to design, build and validate computer-based systems which can assist and support the operations in their various tasks and through this improve the total performance and safety of complex plant operation. Currently, the Halden Project is developing an integrated disturbance handling system for use at nuclear power plants. This paper describes the activities on fault detection and diagnosis within this development project

HAMMLAB 2000 for human factor's studies

International Nuclear Information System (INIS)

Kvalem, J.

1999-01-01

The simulator-based Halden Man-Machine Laboratory (HAMMLAB) has, since its establishment in 1983, been the main vehicle for the human-machine systems research at the OECD Halden Reactor Project. The human factors programme relies upon HAMMLAB for performing experimental studies, but the laboratory is also utilised when evaluating computerised operator support systems, and for experimentation with advanced control room prototypes. The increased focus on experimentation as part of the research programme at the Halden Project, has led to a discussion whether today's laboratory will meet the demands of the future. A pre-project concluded with the need for a new laboratory, with extended simulation capabilities. Based upon these considerations, the HAMMLAB 2000 project was initiated with the goal of making HAMMLAB a global centre of excellence for the study of human-technology interaction in the management and control of industrial processes. This paper will focus on human factors studies to be performed in the new laboratory, and which requirements this will bring upon the laboratory infrastructure and simulation capabilities. The aim of the human factors research at the Halden Project is to provide knowledge which can be used by member organisations to enhance safety and efficiency in the operation of nuclear power plants by utilising research about the capabilities and limitations of the human operator in a control room environment. (author)

REINTRODUCTION OF NOBLE CRAYFISH ASTACUS ASTACUS AFTER CRAYFISH PLAGUE IN NORWAY

Directory of Open Access Journals (Sweden)

TAUGBØL T.

2004-01-01

Full Text Available The Glomma and Halden watercourses in Norway were hit by crayfish plague in 1987 and 1989. Reintroduction of the noble crayfish started in 1989 in the Glomma and in 1995 in the Halden watercourse. Norway has especially good conditions for reintroduction of the native crayfish after crayfish plague, as there is no alien plague-carrying crayfish species in the country. In the Glomma watercourse, approx. 15 000 adult crayfish and 10 000 juveniles have been stocked while in the Halden watercourse the figures are 19 000 adults and 26 500 juveniles. All stocking sites were previously regarded as very good crayfish localities. Four years after stocking, natural recruitment was recorded at all adult crayfish stocking sites in the Glomma watercourse and at most sites in the Halden watercourse. Current crayfish density is, however, much lower than pre-plague densities even at the sites where population development has been in progress for more than 10 years. Extensive post-stocking movements were recorded among adult crayfish. Some sites seemed more suitable for settling, resulting in a great variation in CPUE between the different test-fishing sites. Juveniles seem more appropriate as stocking material if the goal is to re-establish a population in a particular area, due to their stationary behaviour, which seems to remain as they grow larger.

Oxy-fuel combustion as an alternative for increasing lime production in rotary kilns

International Nuclear Information System (INIS)

Granados, D.A.; Chejne, F.; Mejía, J.M.

2015-01-01

Highlights: • A one-dimensional model for oxy-fuel combustion in a rotary kiln was developed. • Flue gas recirculation becomes an important parameter for controlling the process. • Combustion process decreases the flame length making it more dense. • Increases of 12% in raw material with 40% of FGR and conversion of 98% was obtained. - Abstract: The effect of Flue Gas Recirculation (FGR) on the decarbonation process during oxy-fuel combustion in a lime (and cement) rotary kiln is analyzed using an unsteady one-dimensional Eulerian–Lagrangian mathematical model. The model considers gas and limestone as continuous phases and the coal particles as the discrete phase. The model predicts limestone decarbonation, temperature and species distribution of gas and solid phases along the kiln. Simulation results of an air-combustion case are successfully validated with reported experimental data. This model is used to study and to compare the conventional air combustion process with oxy-fuel combustion with FGR ratios between 30% and 80% as controller parameter in this process. Changes in decarbonation process due to energy fluxes by convection and radiation with different FGRs were simulated and analyzed. Simulation results indicate a temperature increase of 20% in the gas and solid phases and a higher decarbonation rate of 40% in relation to the air-combustion case, for a given constant fuel consumption rate. However, for a given temperature, the increase of the CO_2 partial pressure in the oxy-fuel case promotes a reduction of the decarbonation rate. Therefore, there is a compromise between FGR and decarbonation rate, which is analyzed in the present study. Simulation results of the decarbonation step in low FGR cases, compared to air-combustion case, shows that conversion takes place in shorter distances in the kiln, suggesting that the production rate can be increased for existing kilns in oxy-fuel kilns or, equivalently, shorter kilns can be designed for an

Novel use of proton magnetic resonance spectroscopy (1HMRS to non-invasively assess placental metabolism.

Directory of Open Access Journals (Sweden)

Fiona C Denison

Full Text Available Placental insufficiency is a major cause of antepartum stillbirth and fetal growth restriction (FGR. In affected pregnancies, delivery is expedited when the risks of ongoing pregnancy outweigh those of prematurity. Current tests are unable to assess placental function and determine optimal timing for delivery. An accurate, non-invasive test that clearly defines the failing placenta would address a major unmet clinical need. Proton magnetic resonance spectroscopy ((1H MRS can be used to assess the metabolic profile of tissue in-vivo. In FGR pregnancies, a reduction in N-acetylaspartate (NAA/choline ratio and detection of lactate methyl are emerging as biomarkers of impaired neuronal metabolism and fetal hypoxia, respectively. However, fetal brain hypoxia is a late and sometimes fatal event in placental compromise, limiting clinical utility of brain (1H MRS to prevent stillbirth. We hypothesised that abnormal placental (1H MRS may be an earlier biomarker of intrauterine hypoxia, affording the opportunity to optimise timing of delivery in at-risk fetuses.We recruited three women with severe placental insufficiency/FGR and three matched controls. Using a 3T MR system and a combination of phased-array coils, a 20×20×40 mm(1H MRS voxel was selected along the 'long-axis' of the placenta with saturation bands placed around the voxel to prevent contaminant signals. A significant choline peak (choline/lipid ratio 1.35-1.79 was detected in all healthy placentae. In contrast, in pregnancies complicated by FGR, the choline/lipid ratio was ≤0.02 in all placentae, despite preservation of the lipid peak (p<0.001.This novel proof-of-concept study suggests that in severe placental insufficiency/FGR, the observed 60-fold reduction in the choline/lipid ratio by (1H MRS may represent an early biomarker of critical placental insufficiency. Further studies will determine performance of this test and the potential role of 1H-MRS in the in-vivo assessment of

Effects of Assisted Reproduction Technology on Placental Imprinted Gene Expression

Science.gov (United States)

Katagiri, Yukiko; Aoki, Chizu; Tamaki-Ishihara, Yuko; Fukuda, Yusuke; Kitamura, Mamoru; Matsue, Yoichi; So, Akiko; Morita, Mineto

2010-01-01

We used placental tissue to compare the imprinted gene expression of IGF2, H19, KCNQ1OT1, and CDKN1C of singletons conceived via assisted reproduction technology (ART) with that of spontaneously conceived (SC) singletons. Of 989 singletons examined (ART n = 65; SC n = 924), neonatal weight was significantly lower (P < .001) in the ART group than in the SC group, but placental weight showed no significant difference. Gene expression analyzed by real-time PCR was similar for both groups with appropriate-for-date (AFD) birth weight. H19 expression was suppressed in fetal growth retardation (FGR) cases in the ART and SC groups compared with AFD cases (P < .02 and P < .05, resp.). In contrast, CDKN1C expression was suppressed in FGR cases in the ART group (P < .01), while KCNQ1OT1 expression was hyperexpressed in FGR cases in the SC group (P < .05). As imprinted gene expression patterns differed between the ART and SC groups, we speculate that ART modifies epigenetic status even though the possibilities always exist. PMID:20706653

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

Science.gov (United States)

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

2012-05-13

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

Do miRNAs Play a Role in Fetal Growth Restriction? A Fresh Look to a Busy Corner

Directory of Open Access Journals (Sweden)

Benito Chiofalo

2017-01-01

Full Text Available Placenta is the crucial organ for embryo and fetus development and plays a critical role in the development of fetal growth restriction (FGR. There are increasing evidences on the role of microRNAs (miRNAs in a variety of pregnancy-related complications such as preeclampsia and FGR. More than 1880 miRNAs have been reported in humans and most of them are expressed in placenta. In this paper, we aimed to review the current evidence about the topic. According to retrieved data, controversial results about placental expression of miRNAs could be due (at least in part to the different experimental methods used by different groups. Despite the fact that several authors have demonstrated a relatively easy and feasible detection of some miRNAs in maternal whole peripheral blood, costs of these tests should be reduced in order to increase cohorts and have stronger evidence. In this regard, we take the opportunity to solicit future studies on large cohort and adequate statistical power, in order to identify a panel of biomarkers on maternal peripheral blood for early diagnosis of FGR.

New tools and technology for the study of human performance in simulator experiments

International Nuclear Information System (INIS)

Droeivoldsmo, Asgeir

2004-04-01

The Halden Virtual Reality Centre has for the last four years reported a number of experiments in the area of real world application of virtual and augmented reality technology. The insights from these studies have been reviewed and reported as part of a PhD-thesis submitted at the Norwegian University of Science and Technology. This report is based on the thesis and contains a theoretical discussion of how the virtual and augmented reality technology could be used to extend human operator performance in control rooms to include co-operation with plant floor personnel and interaction with not already built equipment. This thesis suggests that new tools and technology can be used for production of relevant data and insights from the study of human performance in simulator and field experiments. It examines some of the theoretical perspectives behind data collection and human performance assessment, and argues for a high resemblance of the real world and use of subject matter expertise in simulator studies. A model is proposed, suggesting that human performance measurement should be tightly coupled to the topic of study and have a close connection to the time line. This coupling requires new techniques for continuous data collection, and eye movement tracking has been identified as a promising basis for this type of measures. One way of improving realism is to create virtual environments allowing for controlling more of the environment surrounding the test subjects. New application areas for virtual environments are discussed for use in control room and field studies. The combination of wearable computing, virtual and augmented (the use of computers to overlay virtual information onto the real world) reality provides many new possibilities to present information to operators. In two experiments, virtual and augmented reality techniques were used to visualise radiation fields for operators in a contaminated nuclear environment. This way the operators could train for

Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978

International Nuclear Information System (INIS)

Crouthamel, C.E.

1978-10-01

This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor

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

Science.gov (United States)

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

2016-03-01

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

Institutt for Energiteknikk - Annual report 1991

International Nuclear Information System (INIS)

1992-01-01

Work at Institutt for Energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The Halden Reactor Project focuses on two main areas: fuel- and materials technology based on experimental work using the Halden reactor; and information technology with emphasis on the development of computer-based systems for controlling and surveilling complex energy plants and industrial processes. The Halden Project represents a collaboration between national safety organizations, national research institutes, power companies and supply industries in 13 OECD countries. Participation by such Eastern European countries as Czechoslovakia is being discussed in concrete terms. At Kjeller irradiation services are based on the JEEP II reactor and the gamma irradiation facility. Neutron irradiation in the reactor produces the radioactive raw materials for the manufacture of radiopharmaceuticals and radiochemicals. Neutron irradiation is also used for the accurate control of conductivity in superpure silicon crystals. The main purpose of IFE's basic research in physics is to utilize neutron beams from the JEEP II reactor for fundamental studies of the physical characteristics of solids and complex liquids. 16 figs

Modelling of the Gadolinium Fuel Test IFA-681 using the BISON Code

Energy Technology Data Exchange (ETDEWEB)

Pastore, Giovanni [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

2016-05-01

In this work, application of Idaho National Laboratory’s fuel performance code BISON to modelling of fuel rods from the Halden IFA-681 gadolinium fuel test is presented. First, an overview is given of BISON models, focusing on UO2/UO2-Gd2O3 fuel and Zircaloy cladding. Then, BISON analyses of selected fuel rods from the IFA-681 test are performed. For the first time in a BISON application to integral fuel rod simulations, the analysis is informed by detailed neutronics calculations in order to accurately capture the radial power profile throughout the fuel, which is strongly affected by the complex evolution of absorber Gd isotopes. In particular, radial power profiles calculated at IFE–Halden Reactor Project with the HELIOS code are used. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project. Some slide have been added as an Appendix to present the newly developed PolyPole-1 algorithm for modeling of intra-granular fission gas release.

Radial power density distribution of MOX fuel rods in the HBWR

International Nuclear Information System (INIS)

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

1999-07-01

Two MOX fuel rods, which ar being fabricated in the Paul Scherrer Institute (PSI), Switzerland in cooperation with the Korea Atomic Energy Research Institute (KAERI), are going to be irradiated in the HBWR (Halden Boiling Water Reactor) from the beginning of 2000 in the framework of OECD Halden Reactor Programme (HRP) together with a reference MOX fuel rod supplied by the BNFL. Since fuel temperature, which is influenced by radial power distribution, is a basic property in analyzing fuel behavior, it is required to consider radial power distribution in the HBWR. A subroutine FACTOR H BWR that calculates radial power density distribution for three MOX fuel rods have been developed subroutine FACTOR H BWR gives good agreement with the physics calculation except slight underprediction in the central part and a little overprediction at the outer part of the pellet. The subroutine will be incorporated into a computer code COSMOS and used to analyze the in-reactor behavior of the three MOX fuel rods during the Halden irradiation test. (author). 5 refs., 3 tabs., 24 figs

Effect of flue gas recirculation on heat transfer in a supercritical circulating fluidized bed combustor

Directory of Open Access Journals (Sweden)

Błaszczuk Artur

2015-09-01

Full Text Available This paper focuses on assessment of the effect of flue gas recirculation (FGR on heat transfer behavior in 1296t/h supercritical coal-fired circulating fluidized bed (CFB combustor. The performance test in supercritical CFB combustor with capacity 966 MWth was performed with the low level of flue gas recirculation rate 6.9% into furnace chamber, for 80% unit load at the bed pressure of 7.7 kPa and the ratio of secondary air to the primary air SA/PA = 0.33. Heat transfer behavior in a supercritical CFB furnace between the active heat transfer surfaces (membrane wall and superheater and bed material has been analyzed for Geldart B particle with Sauter mean diameters of 0.219 and 0.246 mm. Bed material used in the heat transfer experiments had particle density of 2700 kg/m3. A mechanistic heat transfer model based on cluster renewal approach was used in this work. A heat transfer analysis of CFB combustion system with detailed consideration of bed-to-wall heat transfer coefficient distributions along furnace height is investigated. Heat transfer data for FGR test were compared with the data obtained for representative conditions without recycled flue gases back to the furnace through star-up burners.

Assessments of sheath strain and fission gas release data from 20 years of power reactor fuel irradiations

International Nuclear Information System (INIS)

Purdy, P.L.; Manzer, A.M.; Hu, R.H.; Gibb, R.A.; Kohn, E.

1997-01-01

Over the past 20 years, many fuel elements or bundles discharged from Canadian CANDU power reactors have been examined in the AECL hot cells. The post-irradiation examination (PIE) database covers a wide range of operating conditions, from which fuel performance characteristics can be assessed. In the present analysis, a PIE database was compiled representing elements from a total of 129 fuel bundles, of which 26% (34 bundles) were confirmed to have one or more defective elements. This comprehensive database was assessed in terms of measured sheath strain and fission gas release (FGR) for intact elements, in an attempt to identify any changes in these parameters over the history of CANDU reactor operation. Results from this assessment indicate that, for the data that are typical of normal CANDU operating conditions, tensile sheath strain and FGR have remained within 0.5% and 8%, respectively. Those data beyond these ranges are from fuel operated under abnormal conditions, not representative of normal operation, and thus do not indicate a trend toward unexpected fuel behaviour. The distributions of the PIE measurements indicate that maximum expected sheath strains and FGR for normally operated fuel are 0.7% and 13%, respectively. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

NOMAGE4 activities 2011, Part II, Supercritical water loop

DEFF Research Database (Denmark)

Vierstraete, Pierre; Van Nieuwenhove, Rudi; Lauritzen, Bent

The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material...... and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal...

National report from Norway

International Nuclear Information System (INIS)

Haugset, K.

1996-01-01

A wide spectrum of COSSs have been developed within the Halden Project. Major emphasis has been put on assisting the operator in detecting disturbances at an early stage, to avoid development into a more critical situation. Ideas about which systems to develop come either from the Halden research staff itself, or from the participating organizations which consist of licensing organizations, vendors, utilities and research laboratories. The systems are developed up to a prototype level, demonstrated in the HAMMLAB experimental control room. Often, the prototypes are further developed and taken into use by utilities in the participating countries. (author)

IFPE/IFA-508 and 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP

International Nuclear Information System (INIS)

2007-01-01

Description: To measure the integrated response of UO 2 and its cladding to conditions associated with PCI, the Japan Atomic Energy Research Institute carried out a series of experiments in the Halden BWR. The experiment involved two major objectives. The first was to study the influence of rod design parameters on PCI. Diametral gap, wall cladding thickness, SiO 2 additive, and pellet grain size were used as design parameters. The second objective was to study the influence of pre-irradiation (i.e. burnup) on PCI. The maximum burnup attained in the experiment was 23 MWd/kgU. These research results can be applied to current BWR-type fuel rods. The tests were performed between April 1977 and March 1981

Effect of combustion characteristics on wall radiative heat flux in a 100 MWe oxy-coal combustion plant

Energy Technology Data Exchange (ETDEWEB)

Park, S.; Ryu, C. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Chae, T.Y. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Yang, W. [Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Kim, Y.; Lee, S.; Seo, S. [Korea Electric Power Research Institute (KEPRI), Daejeon (Korea, Republic of). Power Generation Lab.

2013-07-01

Oxy-coal combustion exhibits different reaction, flow and heat transfer characteristics from air-coal combustion due to different properties of oxidizer and flue gas composition. This study investigated the wall radiative heat flux (WRHF) of air- and oxy-coal combustion in a simple hexahedral furnace and in a 100 MWe single-wall-fired boiler using computational modeling. The hexahedral furnace had similar operation conditions with the boiler, but the coal combustion was ignored by prescribing the gas properties after complete combustion at the inlet. The concentrations of O{sub 2} in the oxidizers ranging between 26 and 30% and different flue gas recirculation (FGR) methods were considered in the furnace. In the hexahedral furnace, the oxy-coal case with 28% of O{sub 2} and wet FGR had a similar value of T{sub af} with the air-coal combustion case, but its WRHF was 12% higher. The mixed FGR case with about 27% O{sub 2} in the oxidizer exhibited the WRHF similar to the air-coal case. During the actual combustion in the 100 MWe boiler using mixed FGR, the reduced volumetric flow rates in the oxy-coal cases lowered the swirl strength of the burners. This stretched the flames and moved the high temperature region farther to the downstream. Due to this reason, the case with 30% O{sub 2} in the oxidizers achieved a WRHF close to that of air-coal combustion, although its adiabatic flame temperature (T{sub af}) and WHRF predicted in the simplified hexahedral furnace was 103 K and 10% higher, respectively. Therefore, the combustion characteristics and temperature distribution significantly influences the WRHF, which should be assessed to determine the ideal operating conditions of oxy- coal combustion. The choice of the weighted sum of gray gases model (WSGGM) was not critical in the large coal-fired boiler.

Dichorionic twin ultrasound surveillance: sonography every 4 weeks significantly underperforms sonography every 2 weeks: results of the Prospective Multicenter ESPRiT Study.

Science.gov (United States)

Corcoran, Siobhan; Breathnach, Fionnuala; Burke, Gerard; McAuliffe, Fionnuala; Geary, Michael; Daly, Sean; Higgins, John; Hunter, Alyson; Morrison, John J; Higgins, Shane; Mahony, Rhona; Dicker, Patrick; Tully, Elizabeth; Malone, Fergal D

2015-10-01

A 2-week ultrasound scanning schedule for monochorionic twins is endorsed widely. There is a lack of robust data to inform a schedule for the surveillance of dichorionic gestations. We aimed to determine how ultrasound scanning that is performed at 2- or 4-week intervals (or every 4 weeks before 32 weeks' gestation and every 2 weeks thereafter) may impact the prenatal detection of fetal growth restriction (FGR) and ultimately influence timing of delivery. In a consecutive cohort of 789 dichorionic twin pregnancies that were recruited prospectively for the multicenter Evaluation of Sonographic Predictors of Restricted Growth in Twins study, ultrasound determination of fetal growth and interrogation of umbilical and middle cerebral artery Doppler scans were performed every 2 weeks from 24 weeks' gestation until delivery. Complete delivery and perinatal outcome data were recorded for all pregnancies. Where delivery was prompted by FGR, abnormal umbilical artery Doppler examination or poor biophysical profile and in the absence of ruptured membranes, onset of labor, preeclampsia, or antepartum hemorrhage, the delivery was considered "ultrasound-indicated." For ultrasound-indicated deliveries, detection probabilities for FGR/abnormal umbilical artery Doppler scans/poor biophysical were determined according to the interval between examinations, by the suppression if alternate examination data. Among 789 dichorionic twin pregnancies, 66 pairs (8%) had an "ultrasound indicated" delivery. Detection of FGR was reduced from 88-69%, and detection of abnormal umbilical artery Doppler was reduced from 82-62% when a 4-week ultrasound schedule was simulated. Both of these reductions reached statistical significance. There was a nonsignificant trend toward a reduction in the recording of oligohydramnios with a 4-week interval between examinations. This study suggests that the ultrasound surveillance program of every 2 weeks that is recommended currently for monochorionic twins

In situ monitored in-pile creep testing of zirconium alloys

Science.gov (United States)

Kozar, R. W.; Jaworski, A. W.; Webb, T. W.; Smith, R. W.

2014-01-01

The experiments described herein were designed to investigate the detailed irradiation creep behavior of zirconium based alloys in the HALDEN Reactor spectrum. The HALDEN Test Reactor has the unique capability to control both applied stress and temperature independently and externally for each specimen while the specimen is in-reactor and under fast neutron flux. The ability to monitor in situ the creep rates following a stress and temperature change made possible the characterization of creep behavior over a wide stress-strain-rate-temperature design space for two model experimental heats, Zircaloy-2 and Zircaloy-2 + 1 wt%Nb, with only 12 test specimens in a 100-day in-pile creep test program. Zircaloy-2 specimens with and without 1 wt% Nb additions were tested at irradiation temperatures of 561 K and 616 K and stresses ranging from 69 MPa to 455 MPa. Various steady state creep models were evaluated against the experimental results. The irradiation creep model proposed by Nichols that separates creep behavior into low, intermediate, and high stress regimes was the best model for predicting steady-state creep rates. Dislocation-based primary creep, rather than diffusion-based transient irradiation creep, was identified as the mechanism controlling deformation during the transitional period of evolving creep rate following a step change to different test conditions.

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

Science.gov (United States)

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

2018-02-20

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

Design and preparation of scenarios for human factors studies in the HAMMLAB

International Nuclear Information System (INIS)

Moracho, M. J.

1999-01-01

In the preparation of experiments for Human Factors studies, the scenarios play an important role. As a matter of fact, scenario effect is often demonstrated in the analysis of results. An experimental study referred to as Experiments' 97 was conducted in the Halden Man Machine LABoratory. In the design of the scenarios for this experiment, an effort was made for producing scenarios in compliance with the objectives of the study. Before the experiment simulations, scenario analysis was conducted and documented. This report presents some of the lessons learned from these activities. It also identifies main issues to be considered in the scenario characterisation and preparation. Examples of the scenarios' design document produced for the Experiments' 97 as well as examples of measure definitions for OPAS (OPerator Assessment System) and PPAS (Plant Performance System) are included in the appendixes (author) (ml)

Performance of juvenile mojarra supplied with feed containing varying levels of crude protein

Directory of Open Access Journals (Sweden)

Ricardo Henrique Bastos de Souza

2016-04-01

Full Text Available ABSTRACT The growth of the Brazilian aquaculture has stimulated the development of the productive chain of native species, including marine environment. The objective of this study was to evaluate the growth performance of juvenile mojarra fish (Diapterus rhombeus fed diets containing different concentrations of crude protein (32, 36, 40 and 44 g 100 g-1. The 80 juvenile mojarra (7.2±1.5 g were kept in 16 circular tanks (150 L. The study design used was completely randomized with four treatments and four repetitions. The fish were fed four times a day. At the end of the experiment (60 days the final weight, feed intake, weight gain (WG, feed:gain ratio (FGR, protein efficiency rate (PER, energy efficiency rate, specific growth, survival rate and, body composition were evaluated. It was verified significant effect of protein level on the WG, with the best value at the level of 38.20 g 100 g-1 of crude protein. For FGR, the best estimated value occurred with 38.06 g 100 g-1 of crude protein, similar to that reported for the PER (38.91 g 100 g-1. The other performance parameters and body composition were not influenced by crude protein levels. Diet crude protein concentrations between 38.06 and 38.91 g 100 g-1 provide the best performance indices for juvenile mojarra.

1994 report on Task 4.3.1 investigate how training should be planned to ensure efficient utilisation or OSS. Task 4.3.2 evaluate how the introduction of OSS influences requirements to the basic education of the operator

International Nuclear Information System (INIS)

1995-01-01

According to the experience coming from countries where new concepts of nuclear power plants control rooms equipped with integrated system of OSSs have recently been largely tested in experimental centres equipped with powerful real-time plant simulators (e.g. NORWAY - Halden project; Japan) a training in use of the OSS is important, so that the operator can use it in the right situation and in right manner. Experiments, in which mostly experienced active control room operators from operating NPPs took part showed as well, that both the requirements for additional basic education of operators and the planning of training is influenced by a number of different factors like complexity of system structure and interface, system function, capabilities of training centres, operators background, experience and age and others. 6 refs

Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

Energy Technology Data Exchange (ETDEWEB)

Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

2016-08-15

Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

International Nuclear Information System (INIS)

Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

2013-01-01

The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

Investigation of the ramp testing behaviour of fuel pins with different diameters

International Nuclear Information System (INIS)

Pott, G.; Herren, M.; Wigger, B.

1979-09-01

The aim of these experiments was the investigation of the influence of different fuel pin diameter on the ramp testing behaviour. Fuel elements with diameter between 10,75 and 15,6 mm and different cladding thickness had been ramptested in the HBWR (Halden Boiling Water Reactor) after preirradiated in the same facility. Fuel pins with the smallest diameter of 10,75 mm failed. This was indicated by fission gas release measurement. Metallographic examination showed these failure were caused by hydride blisters. A systematic influence of fuel pin diameter and cladding thickness on the ramptesting behaviour was not observed. (orig.) [de

Experience for plant monitoring design in Italian BWR NPP and future trends in man-machine interface

International Nuclear Information System (INIS)

Maestri, F.; Sepielli, M.

1987-01-01

TMI accidental sequence and daily-gained operating experience on italian and abroad NPPs have affected in depth the approach to the design of information presentation to the Control Room staff. It has been cleared that most problems in plant operation arise from a poor and inadequate information system. The main lacks have been identified in the Control Room lay-out and information organization. This has pushed designers both to improve the Control Room environment and to better exploit the computer data processing and data presentation capabilities. The paper deals with the basic criteria for the design and the design review of a computerized system to be inserted in a hybrid Control Room in Italian 981 Mwe BWR-6 NPP, where the concepts outlined above were taken-up from the very beginning. The Control Room keeps conventional instrumentation arranged in a human-factor lay-out, according to post-TMI requirements, and adds a powerful computer-based information system for advanced alarm presentation and plant supervision during both normal and emergency conditions with high data reliability. Colour videounits and operating panels are functionally integrated to create powerful operator work-stations. Emphasis is mostly given on the revision work for video-unit displays and Man-System Communication carried out in cooperation with Halden Reactor Project human factor and plant operation experts. The work peculiarity has been a strong care on the integration between conventional and computerized information presentation, with particular regard to common information and code consistency. (author)

Analysis and comparison of fragrant gene sequence in some rice cultivars

Directory of Open Access Journals (Sweden)

Karami Noushafarin

2016-01-01

Full Text Available It is known that the fragrant trait in rice (Oryza sativa L. is largely controlled by fgr gene on chromosome 8 and it has been specified that the existence of an 8 bp deletion and three single nucleotide polymorphism (SNP in exon 7 is effective on this trait. In this study, sequence alignment analysis of fgr exon7 on chromosome 8 for 11 different fragrant and non-fragrant cultivars revealed that 5 aromatic rice cultivars carried 3 SNPs and 8 bp deletion in exon7 which terminates prematurely at a TAA stop codon. However, 5 of the non-aromatics showed a sequence identical to the published Nipponbare, being non-fragrant Japonica variety sequence. An exception among them was Bejar, which had 8 bp deletion and 3SNPs but it was non-aromatic. Sequencing can determine nucleotide alignment of a gene and give beneficial information about gene function. In silico prediction showed proteins sequences alignment of fgr gene for Khazar and Domsiah genotypes were different. Betaine aldehyde dehydrogenase complete enzyme belongs to Khazar non-fragrant genotype that has complete length and 503 amino acids while non-functional BADH2 enzyme for Domsiah fragrant genotype has 251 amino acids that result in accumulate 2-acetyl-1-pyrroline (2AP and produces aroma in fragrant genotypes.

Association between the high soluble fms-like tyrosine kinase-1 to placental growth factor ratio and adverse outcomes in asymptomatic women with early-onset fetal growth restriction.

Science.gov (United States)

Shinohara, Satoshi; Uchida, Yuzo; Kasai, Mayuko; Sunami, Rei

2017-08-01

To assess whether the high soluble fms-like tyrosine kinase-1 (sFlt-1) to placental growth factor (PlGF) ratio is associated with adverse outcomes (e.g., HELLP syndrome [hemolysis, elevated liver enzymes, and low platelets], severe hypertension uncontrolled by medication, non-reassuring fetal status, placental abruption, pulmonary edema, growth arrest, maternal death, or fetal death) and a shorter duration to delivery in early-onset fetal growth restriction (FGR). Thirty-four women with FGR diagnosed at Women who developed adverse outcomes within a week had a significantly higher sFlt-1/PlGF ratio than did those who did not develop complications. A cutoff value of 86.2 for the sFlt-1/PlGF ratio predicted adverse outcomes, with a sensitivity and specificity of 77.8% and 80.0%, respectively. Moreover, 58.4% of women with an sFlt-1/PlGF ratio ≥86.2 versus 9.1% of those with an sFlt-1/PlGF ratio <86.2 delivered within a week of presentation (p < 0.001). In multivariate analyses, an sFlt-1/PlGF ratio ≥86.2 (adjusted odds ratio 9.52; 95% confidence interval, 1.25-72.8) was associated with adverse maternal and neonatal outcomes. A high sFlt-1/PlGF ratio was associated with adverse outcomes and a shorter duration to delivery in early-onset FGR.

NOMAGE4 activities 2011. Part II, Supercritical water loop

Energy Technology Data Exchange (ETDEWEB)

Vierstraete, P. (Ecole Nationale Superieure des mines, Paris (France)); Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (HRP), Kjeller (Norway)); Lauritzen, B. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy, Roskilde (Denmark))

2012-01-15

The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal of the project is to design a loop allowing material and fuel test studies at significant mass flow with in-core instrumentation and chemistry control possibilities. The present report focusses on the main heat exchanger required for such a loop in the Halden Reactor. The goal of this heat exchanger is to assure a supercritical flow state inside the test section (the core side) and a subcritical flow state inside the pump section. The objective is to design the heat exchanger in order to optimize the efficiency of the heat transfer and to respect several requirements as the room available inside the reactor hall, the maximal total pressure drop allowed and so on. (Author)

Institutt for energiteknikk - Annual Report 1993

International Nuclear Information System (INIS)

1994-01-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the Halden Reactor Project. In 1958, the first Halden Reactor Project Agreement was signed by organisations representing 12 European countries. The project's membership now includes several associated parties in addition to 13 organisations from its 12 NEA members plus one NEA non-member, the Czech Republic. The objectives have evolved from being simply a demonstration of the operation of a boiling heavy-water reactor to becoming a substantial research and development programme covering the domains of a human-machine interaction, fuel behaviour, materials testing, water chemistry, and instrumentation. In 1993, significant progress was achieved in all of the areas addressed by the project, including the re-instrumentation of irradiated fuel rods, fission gas release, irradiation-assisted stress corrosion cracking, a conceptual design of advanced cockpit-type control rooms, analysis of human behaviour, and information processing and presentation. More than 34 years after its reactor's first criticality, the Halden Project continues to provide the scientific community with a wealth of information for the safe and efficient production of nuclear power. The current Project Agreement covers the period from 1993 to 1996

Institutt for Energiteknikk - Annual Report 1994

International Nuclear Information System (INIS)

1995-01-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the Halden Reactor Project. In 1958, the first Halden Reactor Project Agreement was signed by organisations representing 12 European countries. During 1994 France became a full member and associate membership was established with Russia. Accordingly, 16 countries were participating in the Project by the end of the year. The objectives have evolved from being simply a demonstration of the operation of a boiling heavy-water reactor to becoming a substantial research and development programme covering the domains of human-machine interaction, fuel behaviour, materials testing, water chemistry, and instrumentation. In 1994, significant progress was achieved in all of the areas addressed by the project, including the re-instrumentation of irradiated fuel rods, fission gas release, irradiation-assisted stress corrosion cracking, a conceptual design of advanced cockpit-type control rooms, analysis of human behaviour, and information processing and presentation. More than 35 years after its reactor's first critically, the Halden Project continues to provide the scientific community with a wealth of information for the safe and efficient production of nuclear power. The current Project Agreement covers the period from 1993 to 1996

Institutt for Energiteknikk - Annual Report 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the Halden Reactor Project. In 1958, the first Halden Reactor Project Agreement was signed by organisations representing 12 European countries. During 1994 France became a full member and associate membership was established with Russia. Accordingly, 16 countries were participating in the Project by the end of the year. The objectives have evolved from being simply a demonstration of the operation of a boiling heavy-water reactor to becoming a substantial research and development programme covering the domains of human-machine interaction, fuel behaviour, materials testing, water chemistry, and instrumentation. In 1994, significant progress was achieved in all of the areas addressed by the project, including the re-instrumentation of irradiated fuel rods, fission gas release, irradiation-assisted stress corrosion cracking, a conceptual design of advanced cockpit-type control rooms, analysis of human behaviour, and information processing and presentation. More than 35 years after its reactor`s first critically, the Halden Project continues to provide the scientific community with a wealth of information for the safe and efficient production of nuclear power. The current Project Agreement covers the period from 1993 to 1996.

Simulation of integral local tests with high-burnup fuel

International Nuclear Information System (INIS)

Gyori, G.

2011-01-01

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

Increase in SGLT1-mediated transport explains renal glucose reabsorption during genetic and pharmacological SGLT2 inhibition in euglycemia

Science.gov (United States)

Rieg, Timo; Masuda, Takahiro; Gerasimova, Maria; Mayoux, Eric; Platt, Kenneth; Powell, David R.; Thomson, Scott C.; Koepsell, Hermann

2013-01-01

In the kidney, the sodium-glucose cotransporters SGLT2 and SGLT1 are thought to account for >90 and ∼3% of fractional glucose reabsorption (FGR), respectively. However, euglycemic humans treated with an SGLT2 inhibitor maintain an FGR of 40–50%, mimicking values in Sglt2 knockout mice. Here, we show that oral gavage with a selective SGLT2 inhibitor (SGLT2-I) dose dependently increased urinary glucose excretion (UGE) in wild-type (WT) mice. The dose-response curve was shifted leftward and the maximum response doubled in Sglt1 knockout (Sglt1−/−) mice. Treatment in diet with the SGLT2-I for 3 wk maintained 1.5- to 2-fold higher urine glucose/creatinine ratios in Sglt1−/− vs. WT mice, associated with a temporarily greater reduction in blood glucose in Sglt1−/− vs. WT after 24 h (−33 vs. −11%). Subsequent inulin clearance studies under anesthesia revealed free plasma concentrations of the SGLT2-I (corresponding to early proximal concentration) close to the reported IC50 for SGLT2 in mice, which were associated with FGR of 64 ± 2% in WT and 17 ± 2% in Sglt1−/−. Additional intraperitoneal application of the SGLT2-I (maximum effective dose in metabolic cages) increased free plasma concentrations ∼10-fold and reduced FGR to 44 ± 3% in WT and to −1 ± 3% in Sglt1−/−. The absence of renal glucose reabsorption was confirmed in male and female Sglt1/Sglt2 double knockout mice. In conclusion, SGLT2 and SGLT1 account for renal glucose reabsorption in euglycemia, with 97 and 3% being reabsorbed by SGLT2 and SGLT1, respectively. When SGLT2 is fully inhibited by SGLT2-I, the increase in SGLT1-mediated glucose reabsorption explains why only 50–60% of filtered glucose is excreted. PMID:24226519

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

International Nuclear Information System (INIS)

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

2006-01-01

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

NERI Quarterly Progress Report -- April 1 - June 30, 2005 -- Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving PBR Efficiency and Testing Material Compatibility

International Nuclear Information System (INIS)

Chang Oh

2005-01-01

The objective of this research is to improve a helium Brayton cycle and to develop a supercritical carbon dioxide Brayton cycle for the Pebble Bed Reactor (PBR) that can also be applied to the Fast Gas-Cooled Reactor (FGR) and the Very-High-Temperature Gas-Cooled Reactor (VHTR). The proposed supercritical carbon dioxide Brayton cycle will be used to improve the PBR, FGR, and VHTR net plant efficiency. Another objective of this research is to test materials to be used in the power conversion side at supercritical carbon dioxide conditions. Generally, the optimized Brayton cycle and balance of plant (BOP) to be developed from this study can be applied to Generation-IV reactor concepts. Particularly, we are interested in VHTR because it has a good chance of being built in the near future

FY-05 Second Quarter Report On Development of a Supercritical Carbon Dioxide Brayton Cycle: Improving PBR Efficiency and Testing Material Compatibility

International Nuclear Information System (INIS)

Chang Oh

2005-01-01

The objective of this research is to improve a helium Brayton cycle and to develop a supercritical carbon dioxide Brayton cycle for the Pebble Bed Reactor (PBR) that can also be applied to the Fast Gas-Cooled Reactor (FGR) and the Very-High-Temperature Gas-Cooled Reactor (VHTR). The proposed supercritical carbon dioxide Brayton cycle will be used to improve the PBR, FGR, and VHTR net plant efficiency. Another objective of this research is to test materials to be used in the power conversion side at supercritical carbon dioxide conditions. Generally, the optimized Brayton cycle and balance of plant (BOP) to be developed from this study can be applied to Generation-IV reactor concepts. Particularly, we are interested in VHTR because it has a good chance of being built in the near future

An investigation into fuel pulverization with specific reference to high burn-up LOCA

International Nuclear Information System (INIS)

Yagnik, Suresh; Turnbull, James; Noirot, Jean; Walker, Clive; Hallstadius, Lars; Waeckel, N.; Blanpain, P.

2014-01-01

To investigate the phenomenon of high burn-up fuel pellet material potentially disintegrating into powder under a rapid temperature transient, such as in a LOCA-type accident scenario, two independent scoping studies were commissioned. The first was to investigate the effect of hydrostatic restraint pressure on Fission Gas Release (FGR) from small samples of highly irradiated fuel (71 MWd/kgU) during a series of rapid temperature ramps. Experimentally, when the FGR increased rapidly during the temperature transients, the fuel was assumed to be 'pulverized', i.e., fragmented into powder. In the second series of experiments, laser heating of small samples was used to investigate the temperature at which fuel pulverization was initiated. Subsequent to fuel disintegration, there was always a spectrum of particle sizes present. The significance of this observation was recognized in the context of extended burn-up operation in commercial reactors. Based on the observation from these investigations, a fuel fragmentation threshold has been discussed and developed. We conclude that fuel disintegration could be of potential importance in limiting the performance and productive lifetime of nuclear fuel. However, since only fuel closely adjacent to ballooned or ruptured cladding would be released in a LOCA-type transient, expulsion of pulverized fuel from the ruptured fuel rod is not considered a safety issue; cooling of the defected assembly remains possible and there is no issue with respect to local criticality. (author)

OECD Nuclear Energy Agency. 5 activity report. 1976

International Nuclear Information System (INIS)

1977-01-01

The main activities of the Agency are reviewed: nuclear power trends; regulatory aspects of nuclear power; technical developments: Eurochemic, Halden, Dragon, food irradiation; gas-cooled fast reactors, isotopic batteries; nuclear data Centers

Fuel compliance model for pellet-cladding mechanical interaction

International Nuclear Information System (INIS)

Shah, V.N.; Carlson, E.R.

1985-01-01

This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code

Uncertainties in Cancer Risk Coefficients for Environmental Exposure to Radionuclides. An Uncertainty Analysis for Risk Coefficients Reported in Federal Guidance Report No. 13

Energy Technology Data Exchange (ETDEWEB)

Pawel, David [U.S. Environmental Protection Agency; Leggett, Richard Wayne [ORNL; Eckerman, Keith F [ORNL; Nelson, Christopher [U.S. Environmental Protection Agency

2007-01-01

Federal Guidance Report No. 13 (FGR 13) provides risk coefficients for estimation of the risk of cancer due to low-level exposure to each of more than 800 radionuclides. Uncertainties in risk coefficients were quantified in FGR 13 for 33 cases (exposure to each of 11 radionuclides by each of three exposure pathways) on the basis of sensitivity analyses in which various combinations of plausible biokinetic, dosimetric, and radiation risk models were used to generate alternative risk coefficients. The present report updates the uncertainty analysis in FGR 13 for the cases of inhalation and ingestion of radionuclides and expands the analysis to all radionuclides addressed in that report. The analysis indicates that most risk coefficients for inhalation or ingestion of radionuclides are determined within a factor of 5 or less by current information. That is, application of alternate plausible biokinetic and dosimetric models and radiation risk models (based on the linear, no-threshold hypothesis with an adjustment for the dose and dose rate effectiveness factor) is unlikely to change these coefficients by more than a factor of 5. In this analysis the assessed uncertainty in the radiation risk model was found to be the main determinant of the uncertainty category for most risk coefficients, but conclusions concerning the relative contributions of risk and dose models to the total uncertainty in a risk coefficient may depend strongly on the method of assessing uncertainties in the risk model.

The effect of fertility treatment on adverse perinatal outcomes in women aged at least 40 years.

Science.gov (United States)

Harlev, Avi; Walfisch, Asnat; Oran, Eynan; Har-Vardi, Iris; Friger, Michael; Lunenfeld, Eitan; Levitas, Eliahu

2018-01-01

To compare perinatal outcomes between spontaneous conception and assisted reproductive technologies (ART) among patients of advanced maternal age. The present retrospective study included data from singleton pregnancies of women aged at least 40 years who delivered between January 1, 1991, and December 31, 2013, at Soroka University Medical Center, Beer Sheva, Israel. Demographic, obstetric, and perinatal data were compared between pregnancies conceived with ART (in vitro fertilization [IVF] or ovulation induction) and those conceived spontaneously. Multiple regression models were used to define independent predictors of adverse outcomes. A total of 8244 singleton pregnancies were included; 229 (2.8%) following IVF, 86 (1.0%) following ovulation induction, and 7929 (96.2%) were spontaneous. Preterm delivery (P<0.001), fetal growth restriction (FGR) (P<0.001), and cesarean delivery (P<0.001) demonstrated linear associations with the conception mode; the highest rates for each were observed for IVF, with decreased rates for ovulation induction and spontaneous conception. The incidence of gestational diabetes and hypertensive disorders were highest among pregnancies following ART. No association was observed between conception mode and perinatal mortality. Multivariate logistic regression demonstrated that IVF was independently associated with increased odds of preterm delivery (P<0.001) and FGR (P=0.027) compared with spontaneous conception. Among patients of advanced maternal age, ART were independently associated with increased FGR and preterm delivery rates compared with spontaneous pregnancies; perinatal mortality was comparable. © 2017 International Federation of Gynecology and Obstetrics.

Association of maternal and umbilical cord blood leptin concentrations and abnormal color Doppler indices of umbilical artery with fetal growth restriction

Directory of Open Access Journals (Sweden)

Elahe Zareaan

2017-08-01

Full Text Available Background: Fetal growth restriction (FGR is a condition with heterogeneous pathophysiology which characterized by fetal weight less than the tenth percentile for gestational age. Several factors have impact on maternal, placental and fetal due to growth restriction. Objective: The aim of this study was to investigate the relationship between levels of leptin in the cord, and serum leptin of mothers also abnormal color Doppler indices of umbilical artery with fetal growth restriction. Materials and Methods: This is a cross sectional study conducted in Isfahan, Iran, 2015-2016. We recruited 40 women with singleton pregnancies complicated by fetal growth restriction (Group I and 40 pregnant women with normal fetal growth (Group II with matched age. Maternal serum and umbilical artery leptin levels were determined with Enzyme-Linked immunosorben method. Also, color Doppler ultrasound of umbilical artery was performed. Results: Mean maternal and fetal leptin levels were lower in the FGR group compared to the normal group (36.58±(20.99 and 7.42 ±(4.08vs. 47.32±(22.50 and 30.49±(14.50 respectively. Also, mean fetal leptin level was lower in the group with abnormal color Doppler sonographic indices compared to the normal group (7. 40 ±(4.10vs 27.06±(15.80, respectively. Conclusion: This study indicated that maternal and fetal leptin levels are correlated with FGR originating from damaged placental function; also fetal leptin level can indicate changes in color Doppler sonographic indices.

Institute for Energy Technology -Annual report 1996

International Nuclear Information System (INIS)

1997-01-01

Research at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the OECD Halden Reactor Project. 19 participating countries and about 100 organisations is involved in the project. The Project is operated by a staff of 280 persons. In the autumn of 1996 the participating organizations reached agreement to continue their research collaboration for a further 3-year period (1997 to 1999). An extensive experimental program was carried out in 1996 using the Halden reactor (HBWR), partly for the joint international program, and partly for contract work for member countries. The main aim of this work is to improve the safety and reliability of existing nuclear power plants. The experimental equipment in the Halden reactor makes it ideal for simulating various operating conditions in different types of rectors. Processes such as corrosion in fuel cladding materials and fracture propagation in irradiated materials under the influence of additives in the coolant water can be studied. In an on-going study, fuel of Russian origin is being compared with modern western fuel. The results, being the first of their kind that are openly available, form an important bases for safety assessments of Russian VVER reactors. The man-machine laboratory is used to study how new technologies influence the operator and to develop computer based systems for improving the safety and accessibility of complex processes

Development of high performance cladding materials

International Nuclear Information System (INIS)

Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

2010-04-01

The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

Institute for Energy Technology -Annual report 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

Research at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the OECD Halden Reactor Project. 19 participating countries and about 100 organisations is involved in the project. The Project is operated by a staff of 280 persons. In the autumn of 1996 the participating organizations reached agreement to continue their research collaboration for a further 3-year period (1997 to 1999). An extensive experimental program was carried out in 1996 using the Halden reactor (HBWR), partly for the joint international program, and partly for contract work for member countries. The main aim of this work is to improve the safety and reliability of existing nuclear power plants. The experimental equipment in the Halden reactor makes it ideal for simulating various operating conditions in different types of rectors. Processes such as corrosion in fuel cladding materials and fracture propagation in irradiated materials under the influence of additives in the coolant water can be studied. In an on-going study, fuel of Russian origin is being compared with modern western fuel. The results, being the first of their kind that are openly available, form an important bases for safety assessments of Russian VVER reactors. The man-machine laboratory is used to study how new technologies influence the operator and to develop computer based systems for improving the safety and accessibility of complex processes.

Correlation between Doppler flow patterns in growth-restricted fetuses and neonatal circulation

NARCIS (Netherlands)

Tanis, J. C.; Boelen, M. R.; Schmitz, D. M.; Casarella, L.; van der Laan, M. E.; Bos, A. F.; Bilardo, C. M.

Objectives To investigate whether prenatal Doppler parameters in growth-restricted fetuses are correlated with neonatal circulatory changes. Methods In 43 cases of suspected fetal growth restriction (FGR), serial Doppler measurements of umbilical artery (UA) and middle cerebral artery (MCA)

Electrochemical Study of Corrosion Phenomena in Zirconium Alloys

National Research Council Canada - National Science Library

Treeman, Nicole M

2005-01-01

..., has become a potentially life-limiting issue for BWR fuel. Recent results from experimentation at MIT, Halden, and Studvik suggest that a galvanic coupling drives the phenomenon between the cladding and the adjacent material...

Neutron Stars : Magnetism vs Gravity

Indian Academy of Sciences (India)

however, in the magnetosphere, electromagnetic forces dominate over gravity : Fgr = mg ~ 10-18 Newton ; Fem = e V B ~ 10-5 Newton; (for a single electron of mass m and charge e ) ; Hence, the electromagnetic force is 1013 times stronger than gravity !!

Simulating fuel behavior under transient conditions using FRAPTRAN and uncertainty analysis using Dakota

International Nuclear Information System (INIS)

Gomes, Daniel S.; Teixeira, Antonio S.

2017-01-01

Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)

Simulating fuel behavior under transient conditions using FRAPTRAN and uncertainty analysis using Dakota

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Teixeira, Antonio S., E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)

OECD Nuclear Energy Agency. 3. Activity report, 1974

International Nuclear Information System (INIS)

1975-01-01

The main activities of the Agency are reviewed: study of nuclear power trends; regulatory aspects of nuclear power; technical developments: Eurochemic, Halden, Dragon, food irradiation, gas-cooled fast reactors, direct conversion, isotopic batteries; nuclear energy information

Early, Incomplete, or Preclinical Autoimmune Systemic Rheumatic Diseases and Pregnancy Outcome.

Science.gov (United States)

Spinillo, Arsenio; Beneventi, Fausta; Locatelli, Elena; Ramoni, Vèronique; Caporali, Roberto; Alpini, Claudia; Albonico, Giulia; Cavagnoli, Chiara; Montecucco, Carlomaurizio

2016-10-01

To evaluate the impact of preclinical systemic autoimmune rheumatic disorders on pregnancy outcome. In this longitudinal cohort study, patients were enrolled during the first trimester of pregnancy if they reported having had connective tissue disorder symptoms, were found to be positive for circulating autoantibodies, and on clinical evaluation were judged to have a preclinical or incomplete rheumatic disorder. The incidence of fetal growth restriction (FGR), preeclampsia, and adverse pregnancy outcomes in patients with preclinical rheumatic disorders was compared with that in selected controls, after adjustment for confounders by penalized logistic regression. Odds ratios (ORs) and 95% confidence intervals (95% CIs) were calculated. Of 5,232 women screened, 150 (2.9%) were initially diagnosed as having a suspected rheumatic disorder. After a mean ± SD postpartum follow-up of 16.7 ± 5.5 months, 64 of these women (42.7%) had no clinically apparent rheumatic disease and 86 (57.3%) had persistent symptoms and positive autoantibody results, including 10 (6.7%) who developed a definitive rheumatic disease. The incidences of preeclampsia/FGR and of small for gestational age (SGA) infants were 5.1% (23 of 450) and 9.3% (42 of 450), respectively, among controls, 12.5% (8 of 640) (OR 2.7 [95% CI 1.1-6.4]) and 18.8% (12 of 64) (OR 2.2 [95% CI 1.1-4.5]), respectively, among women with no clinically apparent disease, and 16.3% (14 of 86) (OR 3.8 [95% CI 1.9-7.7]) and 18.6% (16 of 86) (OR 2.3 [95% CI 1.2-4.3]), respectively, among those with persisting symptoms at follow-up. Mean ± SD umbilical artery Doppler pulsatility indices were higher among women with no clinically apparent disease (0.95 ± 0.2) and those with persisting symptoms (0.96 ± 0.21) than in controls (0.89 ± 0.12) (P = 0.01 and P rheumatic disorders were associated with an increased risk of FGR/preeclampsia and SGA. The impact of these findings and their utility in screening

How to monitor pregnancies complicated by fetal growth restriction and delivery before 32 weeks : post-hoc analysis of TRUFFLE study

NARCIS (Netherlands)

Ganzevoort, W.; Mensing van Charante, N.; Thilaganathan, B.; Prefumo, Federico; Arabin, B.; Bilardo, Caterina M.; Brezinka, C.; Derks, J. B.; Diemert, A.; Duvekot, Johannes J.; Ferrazzi, E.; Frusca, T.; Hecher, K.; Marlow, N.; Martinelli, P.; Ostermayer, E.; Papageorghiou, Aris T.; Schlembach, D.; Schneider, K. T M; Todros, T.; Valcamonico, A.; Visser, G. H.A.; van Wassenaer-Leemhuis, A.; Lees, Christoph C.; Wolf, H.; Aktas, Ayse; Borgione, Silvia; Chaoui, Rabih; Cornette, Jerome M J; Diehl, Thilo; van Eyck, J.; Fratelli, Nicola; van Haastert, I. C.; Lobmaier, Silvia; Lopriore, E.; Missfelder-Lobos, Hannah; Mansi, Giuseppina; Martelli, Paola; Maso, Gianpaolo; Maurer-Fellbaum, Ute; Mulder-De Tollenaer, Susanne; Napolitano, Raffaele; Oberto, Manuela; Oepkes, D.; Ogge, Giovanna; van der Post, Joris A. M.; Preston, Lucy; Raimondi, Francesco; Rattue, H.; Reiss, Irwin K M; Scheepers, L. S.; Skabar, Aldo; Spaanderman, M.; Weisglas-Kuperus, N.; Zimmermann, Andrea

2017-01-01

Objectives: In the recent TRUFFLE study, it appeared that, in pregnancies complicated by fetal growth restriction (FGR) between 26 and 32 weeks' gestation, monitoring of the fetal ductus venosus (DV) waveform combined with computed cardiotocography (CTG) to determine timing of delivery increased the

Interstitial Pregnancy Resulting in a Viable Infant Coexistent with Massive Perivillous Fibrin Deposition: A Case Report and Literature Review

Directory of Open Access Journals (Sweden)

Yusuke Tanaka

2014-05-01

Full Text Available Objective - The objective of this report is to describe a rare case of interstitial pregnancy ultimately resulting in a viable infant coexistent with massive perivillous fibrin deposition (MPFD. Study Design - This study is a case report and literature review. Results - A 35-year-old female patient underwent cesarean section at 32 weeks of gestation due to fetal growth restriction (FGR and breech presentation. During the operation, a diagnosis of interstitial pregnancy was established. There was no evidence of placental separation. We decided to complete surgery without removal of the placenta and waited until the placenta delivered spontaneously. The conservative management was successful, and the patient was discharged on postoperative day 13. The pathologic examination showed MPFD. Conclusion - If interstitial pregnancies are not diagnosed at an early gestational age, it can result in a viable fetus, but such pregnancies may be associated with FGR or placenta accreta.

Complexity rating of abnormal events and operator performance

International Nuclear Information System (INIS)

Oeivind Braarud, Per

1998-01-01

The complexity of the work situation during abnormal situations is a major topic in a discussion of safety aspects of Nuclear Power plants. An understanding of complexity and its impact on operator performance in abnormal situations is important. One way to enhance understanding is to look at the dimensions that constitute complexity for NPP operators, and how those dimensions can be measured. A further step is to study how dimensions of complexity of the event are related to performance of operators. One aspect of complexity is the operator 's subjective experience of given difficulties of the event. Another related aspect of complexity is subject matter experts ratings of the complexity of the event. A definition and a measure of this part of complexity are being investigated at the OECD Halden Reactor Project in Norway. This paper focus on the results from a study of simulated scenarios carried out in the Halden Man-Machine Laboratory, which is a full scope PWR simulator. Six crews of two licensed operators each performed in 16 scenarios (simulated events). Before the experiment subject matter experts rated the complexity of the scenarios, using a Complexity Profiling Questionnaire. The Complexity Profiling Questionnaire contains eight previously identified dimensions associated with complexity. After completing the scenarios the operators received a questionnaire containing 39 questions about perceived complexity. This questionnaire was used for development of a measure of subjective complexity. The results from the study indicated that Process experts' rating of scenario complexity, using the Complexity Profiling Questionnaire, were able to predict crew performance quite well. The results further indicated that a measure of subjective complexity could be developed that was related to crew performance. Subjective complexity was found to be related to subjective work load. (author)

Evidence for accretion of fine-grained rims in a turbulent nebula for CM Murchison

Science.gov (United States)

Hanna, Romy D.; Ketcham, Richard A.

2018-01-01

We use X-ray computed tomography (XCT) to examine the 3D morphology and spatial relationship of fine-grained rims (FGRs) of Type I chondrules in the CM carbonaceous chondrite Murchison to investigate the formation setting (nebular vs. parent body) of the FGRs. We quantify the sizes, shapes, and orientations of the chondrules and FGRs and develop a new algorithm to examine the 3D variation of FGR thickness around each chondrule. We find that the average proportion of chondrule volume contained in the rim for Murchison chondrules is 35.9%. The FGR volume in relation to the interior chondrule radius is well described by a power law function as proposed for accretion of FGRs in a weakly turbulent nebula by Cuzzi (2004). The power law exponent indicates that the rimmed chondrules behaved as Stokes number Stη > 1 nebular particles in Kolmogorov η scale turbulence. FGR composition as inferred from XCT number appears essentially uniform across interior chondrule types and compositions, making formation by chondrule alteration unlikely. We determine that the FGRs were compressed by the impact event(s) that deformed Murchison (Hanna et al., 2015), resulting in rims that are thicker in the plane of foliation but that still preserve their nebular morphological signature. Finally, we propose that the irregular shape of some chondrules in Murchison is a primary feature resulting from chondrule formation and that chondrules with a high degree of surface roughness accreted a relatively larger amount of nebular dust compared to smoother chondrules.

Melatonin rescues cardiovascular dysfunction during hypoxic development in the chick embryo.

Science.gov (United States)

Itani, Nozomi; Skeffington, Katie L; Beck, Christian; Niu, Youguo; Giussani, Dino A

2016-01-01

There is a search for rescue therapy against fetal origins of cardiovascular disease in pregnancy complicated by chronic fetal hypoxia, particularly following clinical diagnosis of fetal growth restriction (FGR). Melatonin protects the placenta in adverse pregnancy; however, whether melatonin protects the fetal heart and vasculature in hypoxic pregnancy independent of effects on the placenta is unknown. Whether melatonin can rescue fetal cardiovascular dysfunction when treatment commences following FGR diagnosis is also unknown. We isolated the effects of melatonin on the developing cardiovascular system of the chick embryo during hypoxic incubation. We tested the hypothesis that melatonin directly protects the fetal cardiovascular system in adverse development and that it can rescue dysfunction following FGR diagnosis. Chick embryos were incubated under normoxia or hypoxia (14% O2) from day 1 ± melatonin treatment (1 mg/kg/day) from day 13 of incubation (term ~21 days). Melatonin in hypoxic chick embryos rescued cardiac systolic dysfunction, impaired cardiac contractility and relaxability, increased cardiac sympathetic dominance, and endothelial dysfunction in peripheral circulations. The mechanisms involved included reduced oxidative stress, enhanced antioxidant capacity and restored vascular endothelial growth factor expression, and NO bioavailability. Melatonin treatment of the chick embryo starting at day 13 of incubation, equivalent to ca. 25 wk of gestation in human pregnancy, rescues early origins of cardiovascular dysfunction during hypoxic development. Melatonin may be a suitable antioxidant candidate for translation to human therapy to protect the fetal cardiovascular system in adverse pregnancy. © 2015 The Authors. Journal of Pineal Research. Published by John Wiley & Sons Ltd.

Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code

International Nuclear Information System (INIS)

Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.

2002-01-01

The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)

Longitudinal study of computerized cardiotocography in early fetal growth restriction

NARCIS (Netherlands)

Wolf, H.; Arabin, B.; Lees, Christoph C.; Oepkes, D.; Prefumo, Federico; Thilaganathan, B.; Todros, T.; Visser, G.H.A.; Bilardo, Caterina M.; Derks, J. B.; Diemert, A.; Duvekot, Johannes J.; Ferrazzi, E.; Frusca, T.; Hecher, K.; Marlow, N.; Martinelli, P.; Ostermayer, E.; Papageorghiou, Aris T.; Scheepers, Hubertina C. J.; Schlembach, D.; Schneider, K. T M; Valcamonico, A.; van Wassenaer-Leemhuis, A.; Ganzevoort, W.; Aktas, Ayse; Borgione, Silvia; Brezinka, Christoph; Calvert, Sandra; Chaoui, Rabih; Cornette, Jerome M J; Diehl, Thilo; van Eyck, Jim; Fratelli, Nicola; van Haastert, Inge Lot; Johnson, Samantha; Lobmaier, Silvia; Lopriore, Enrico; Mansi, Giuseppina; Missfelder-Lobos, Hannah; Martelli, Paola; Maso, Gianpaolo; Maurer-Fellbaum, Ute; Van Charante, Nico Mensing; De Tollenaer, Susanne Mulder; Moore, Tamanna; Napolitano, Raffaele; Oberto, Manuela; Ogge, Giovanna; van der Post, Joris Am; Preston, Lucy; Raimondi, Francesco; Reiss, Irwin K M; Rigano, Serena; Schuit, Ewoud; Skabar, Aldo; Spaanderman, Marc E.; Weisglas-Kuperus, Nynke; Zimmermann, Andrea

2017-01-01

Objectives: To explore whether, in early fetal growth restriction (FGR), the longitudinal pattern of fetal heart rate (FHR) short-term variation (STV) can be used to identify imminent fetal distress and whether abnormalities of FHR recordings are associated with 2-year infant outcome. Methods: The

Robotics and remote maintenance concepts for fusion machines

International Nuclear Information System (INIS)

1989-02-01

Descriptions of operation and maintenance of current tokamaks (TFTR, JET, JT-60) is discussed in the context of radioactivation resulting from thermonuclear reactions. Plans for future devices (NET, CIT, FGR) with respect to remote handling, maintenance, measurements, and robotics are discussed. Refs, figs and tabs

Defining the residual risk of adverse perinatal outcome in growth restricted fetuses with normal umbilical artery blood flow.

LENUS (Irish Health Repository)

O'Dwyer, Vicky

2014-07-25

To determine the cause of adverse perinatal outcome in fetal growth restriction(FGR) where umbilical artery Doppler(UA) was normal, as identified from the Prospective Observational Trial to Optimize Pediatric Health(PORTO). We compared cases of adverse outcome where UA Doppler was normal and abnormal.

Thermal-mechanical properties of cracked UO2 pellets

International Nuclear Information System (INIS)

Williford, R.E.; Mohr, C.L.; Lanning, D.D.

1980-11-01

A series of experiments (IFA-431, 432, 513, and 527) sponsored by the Fuel Behavior Research Branch of the USNRC are being irradiated in the Halden Boiling Water Reactor to better define LWR fuel behavior over the normal operating range of power reactor fuel rods. One fuel behavior variable of interest is the thermally induced cracking of UO 2 fuel pellets. The effects of pellet cracking on the effective thermal conductivity and elastic moduli for the fragmented fuel were found to be primarily dependent on the free area in the r, theta plane of the fuel rod. The free area is defined as the area within the cladding inner surface that is not occupied by the fuel fragments themselves

A review of fission gas release data within the Nea/IAEA IFPE database

International Nuclear Information System (INIS)

Turnbull, J.A.; Menut, P.; Sartori, E.

2002-01-01

The paper describes the International Fuel Performance Experimental (IFPE) database on nuclear fuel performance. The aim of the project is to provide a comprehensive and well-qualified database on Zr clad UO 2 fuel for model development and code validation in the public domain. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in material testing reactors. To date, the database contains some 380 individual cases, the majority of which provide data on FGR either from in-pile pressure measurements or PIE techniques including puncturing, electron probe microanalysis (EPMA) and X-ray fluorescence (XRF) measurements. The paper outlines parameters affecting fission gas release and highlights individual datasets addressing these issues. (authors)

Performance of Bruce natural UO2 fuel irradiated to extended burnups

International Nuclear Information System (INIS)

Zhou, Y.N.; Floyd, M.R.; Ryz, M.A.

1995-11-01

Bruce-type bundles XY, AAH and GF were successfully irradiated in the NRU reactor at Chalk River Laboratories to outer-element burnups of 570-900 MWh/kgU. These bundles were of the Bruce Nuclear Generating Station (NGS)-A 'first-charge' design that contained gas plenums in the outer elements. The maximum outer-element linear powers were 33-37 kW/m. Post-irradiation examination of these bundles confirmed that all the elements were intact. Bundles XY and AAH, irradiated to outer-element burnups of 570-700 MWh/kgU, experienced low fission-gas release (FGR) ( 500 MWh/kgU (equivalent to bundle-average 450 MWh/kgU) when maximum outer-element linear powers are > 50 kW/m. The analysis in this paper suggests that CANDU 37-element fuel can be successfully irradiated (low-FGR/defect-free) to burnups of at least 700 MWh/kgU, provided maximum power do not exceed 40 kW/m. (author). 5 refs., 1 tab., 8 figs

Growth throughout childhood of children born growth restricted

NARCIS (Netherlands)

Beukers, Fenny; Rotteveel, Joost; van Weissenbruch, Mirjam M.; Ganzevoort, Wessel; van Goudoever, Johannes B.; van Wassenaer-Leemhuis, Aleid G.

2017-01-01

Many studies that examine growth in growth-restricted children at birth do not discriminate between fetal growth restriction (FGR) and small for gestational age (SGA). These terms however are not synonymous. In SGA, stunting and increased weight gain have been reported. We do not know if this holds

PEANO: advancements in 1998-99

International Nuclear Information System (INIS)

Fantoni, Paolo F.; Hoffmann, Mario; Lipcsei, Sandor; Roverso, Davide

1999-04-01

PEANO is the signal validation toolbox developed at Halden in the years 1994-99. It is based on a neuro-fuzzy model that is able to assess on-line the confidence of the validation task. The first beta release has been presented at the Enlarged Halden Programme Group Meeting at Lillehammer, March 1998. Since then, several tests have been performed and the feedback and lessons learned by those tests have been used to improve the system significantly. The main enhancements reported here include; Improvements in the training methods, redesign of the user interface and the database management system, the development of wavelet based denoising filters for the training datasets, the development of an interactive designer of IIR digital filters for real-time signal noise suppression, PEANO runs on Windows NT 4.0 platforms and requires the use of MATLAB 5.0 by Math Works Inc (author) (ml)

Ecological interface design for turbine secondary systems in a nuclear power plant : effects on operator situation awareness

International Nuclear Information System (INIS)

Kwok, J.

2007-01-01

Investigations into past accidents at nuclear power generating facilities such as that of Three Mile Island have identified human factors as one of the foremost critical aspects in plant safety. Errors resulting from limitations in human information processing are of particular concern for human-machine interfaces (HMI) in plant control rooms. This project examines the application of Ecological Interface Design (EID) in HMI information displays and the effects on operator situation awareness (SA) for turbine secondary systems based on the Swedish Forsmark 3 boiling-water reactor nuclear power plant. A work domain analysis was performed on the turbine secondary systems yielding part-whole decomposition and abstraction hierarchy models. Information display requirements were subsequently extracted from the models. The resulting EID information displays were implemented in a full-scope simulator and evaluated with six licensed operating crews from the Forsmark 3 plant. Three measures were used to examine SA: self-rated bias, Halden Open Probe Elicitation (HOPE), and Situation Awareness Control Room Inventory (SACRI). The data analysis revealed that operators achieved moderate to good SA; operators unfamiliar with EID information displays were able to develop and maintain comparable levels of SA to operators using traditional forms of single sensor-single indicator (SS-SI) information displays. With sufficient training and experience, operator SA is expected to benefit from the knowledge-based visual elements in the EID information displays. This project was researched in conjunction with the Cognitive Engineering Laboratory at the University of Toronto and the Institute for Energy Technology (IFE) in Halden, Norway. (author)

Annual report 1983

International Nuclear Information System (INIS)

1983-01-01

Nuclear power, isotope technology and basic research in physics comprises the nuclear activities of Institute for Energy Technology (IFE), Norway. These activities represented in 1983 about 60% of the gross turnover and use about 50% of the state grant. IFE's nuclear power work is now mainly connected with the Halden reactor, where international collaboration through the OECD Halden Reactor Project enables Norway to carry out advanced research in central areas of nuclear power technology. The present programme in Halden is concentrated on fuel and safety research, as well as computerbased methods for control and surveillance of reactor installations. An important part of the reactor safety work at IFE is concentrated on Nordic collaboration projects under the direction of the Nordic Coordination Committee for Atomic Energy (NKA). The isotope laboratories activity includes the provision of isotopes, irradiation technology assignements and the development of new products and application methods in the field. Activities within basic research in physics are largely based on the use of neutron beams from the JEEP II reactor for research into the structure and dynamics of solid and liquid materials. The research activity has a broad contact network with Norwegian and foreign research groups. During 1983 IFE continued its countrywide monitoring of radioactive fallout which it began in 1982 as an assignment for the Norwegian Defence Research Establishment. The Committee of IFE's Founders' report on IFE's place within Norwegian research was handled by the authorities and interested parties during 1983. This report is seen as an important foundation for IFE's further development and effort. (RF)

The impact of unrecognized autoimmune rheumatic diseases on the incidence of preeclampsia and fetal growth restriction: a longitudinal cohort study.

Science.gov (United States)

Spinillo, Arsenio; Beneventi, Fausta; Locatelli, Elena; Ramoni, Vèronique; Caporali, Roberto; Alpini, Claudia; Albonico, Giulia; Cavagnoli, Chiara; Montecucco, Carlomaurizio

2016-10-18

The burden of pregnancy complications associated with well defined, already established systemic rheumatic diseases preexisting pregnancy such as rheumatoid arthritis, systemic lupus erythematosus or scleroderma is well known. Systemic rheumatic diseases are characterized by a long natural history with few symptoms, an undifferentiated picture or a remitting course making difficult a timely diagnosis. It has been suggested that screening measures for these diseases could be useful but the impact of unrecognized systemic rheumatic disorders on pregnancy outcome is unknown. The objective of the study was to evaluate the impact of previously unrecognized systemic autoimmune rheumatic on the incidence of preeclampsia and fetal growth restriction (FGR). A longitudinal cohort-study with enrolment during the first trimester of pregnancy of women attending routine antenatal care using a two-step approach with a self-reported questionnaire, autoantibody detection and clinical evaluation of antibody-positive subjects. The incidence of FGR and preeclampsia in subjects with newly diagnosed rheumatic diseases was compared to that of selected negative controls adjusting for potential confounders by logistic regression analysis. The prevalence of previously unrecognized systemic rheumatic diseases was 0.4 % for rheumatoid arthritis (19/5232), 0.25 % (13/5232) for systemic lupus erythematosus, 0.31 % (16/5232) for Sjögren's syndrome, 0.3 % for primary antiphospholipid syndrome (14/5232) and 0.11 % (6/5232) for other miscellaneous diseases. Undifferentiated connective tissue disease was diagnosed in an additional 131 subjects (2.5 %). The incidence of either FGR or preeclampsia was 6.1 % (36/594) among controls and 25.3 % (50/198) in subjects with unrecognized rheumatic diseases (excess incidence = 3.9 % (95 % CI = 2.6-9.6) or 34 % (95 % CI = 22-44) of all cases of FGR/preeclampsia). The incidence of small for gestational age infant (SGA) was higher among

Communication Profile of Primary School-Aged Children with Foetal Growth Restriction

Science.gov (United States)

Partanen, Lea Aulikki; Olsén, Päivi; Mäkikallio, Kaarin; Korkalainen, Noora; Heikkinen, Hanna; Heikkinen, Minna; Yliherva, Anneli

2017-01-01

Foetal growth restriction is associated with problems in neurocognitive development. In the present study, prospectively collected cohorts of foetal growth restricted (FGR) and appropriate for gestational age grown (AGA) children were examined at early school-age by using the Children's Communication Checklist-2 (CCC-2) to test the hypothesis that…

Implementation of Software Tools for Hybrid Control Rooms in the Human Systems Simulation Laboratory

International Nuclear Information System (INIS)

Jokstad, Håkon; Berntsson, Olof; McDonald, Robert; Boring, Ronald; Hallbert, Bruce; Fitzgerald, Kirk

2014-01-01

The Institute for Energy Technology (IFE) and Idaho National Laboratory have designed, implemented, tested and installed a functioning prototype of a set of large screen overview and procedure support displays for the Generic Pressurized Water Reactor (GPWR) simulator in the U.S. Department of Energy's Human Systems Simulation Laboratory. The overview display is based on IFE's extensive experiences with large screen overview displays in the Halden Man-Machine Laboratory (HAMMLAB), and presents the main control room indicators on a combined three-screen display. The procedure support displays are designed and implemented to provide a compact but still comprehensive overview of the relevant process measurements and indicators to support operators' good situational awareness during the performance of various types of procedures and plant conditions.

FCI: remedy development for the fuel performance improvement program

International Nuclear Information System (INIS)

Buckman, F.W.; Crouthamel, C.E.; Freshley, M.D.

1979-01-01

Out-of-reactor experiments and irradiations are being utilized to develop and demonstrate the efficacy of specific advanced fuel designs to improve FCI behavior. The advanced light water reactor fuel designs being evaluated combine annular pellets, graphite coating on the inner surface of the cladding, and helium pressurization. A sphere-pac fuel design is also being developed. Characterization of the graphite coatings includes studies of composition, application methods, thickness control, moisture control, thermal conductivity, compatibility with the zircaloy cladding, strain-to-failure, and friction and wear characteristics. Rods of the different fuel designs, as well as reference rods, are being irradiated in the Halden Boiling Water Reactor and the Big Rock Point Reactor to accumulate burnup prior to ramping tests

Implementation of Software Tools for Hybrid Control Rooms in the Human Systems Simulation Laboratory

Energy Technology Data Exchange (ETDEWEB)

Jokstad, Håkon [Halden Reactor Project, Halden (Norway); Berntsson, Olof [Halden Reactor Project, Halden (Norway); McDonald, Robert [Halden Reactor Project, Halden (Norway); Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fitzgerald, Kirk [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-11-01

The Institute for Energy Technology (IFE) and Idaho National Laboratory have designed, implemented, tested and installed a functioning prototype of a set of large screen overview and procedure support displays for the Generic Pressurized Water Reactor (GPWR) simulator in the U.S. Department of Energy’s Human Systems Simulation Laboratory. The overview display is based on IFE’s extensive experiences with large screen overview displays in the Halden Man-Machine Laboratory (HAMMLAB), and presents the main control room indicators on a combined three-screen display. The procedure support displays are designed and implemented to provide a compact but still comprehensive overview of the relevant process measurements and indicators to support operators' good situational awareness during the performance of various types of procedures and plant conditions.

Abnormal umbilical artery Doppler velocimetry and placental ...

African Journals Online (AJOL)

The study was prospective and conducted in a low-income setting. A total of 130 non-anomalous singleton FGR pregnancies (≥24 weeks) were included in the study. All pregnancies were confirmed to be small for gestational age (SGA) after the birth of the neonate. The placental lesions and neonatal outcomes were ...

Summary of the findings of the FUMEX programme

International Nuclear Information System (INIS)

Chantoin, P.; Wisenack, W.

1997-01-01

Description of fuel behavior during normal, transient and accident conditions has always represented a most challenging and important problem. The ultimate goal is a description of fuel behavior in all conditions to derive safety rules, improved design and economics. The FUMEX programme, promoted by the Agency with the support of the Halden project, is the second co-ordinated research programme in this area. The first step of FUMEX was to conduct a blind exercise which was carried out using 19 computer codes on 10 experimental rods irradiated in the Halden reactor. The results of this exercise are reported here and show an important improvement of modelling tools since 1984, especially due to the development of national and international parametric studies. However, shortcomings still exist and improvement in the evaluation of PCI, fuel temperature and fission gas release, especially during ramp, is not very well understood. The fuel clad gap modelling also requires further improvement. (author). 7 refs, 7 figs, 4 tabs

Summary of the findings of the FUMEX programme

Energy Technology Data Exchange (ETDEWEB)

Chantoin, P [International Atomic Energy Agency, Vienna (Austria); Wisenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

1997-08-01

Description of fuel behavior during normal, transient and accident conditions has always represented a most challenging and important problem. The ultimate goal is a description of fuel behavior in all conditions to derive safety rules, improved design and economics. The FUMEX programme, promoted by the Agency with the support of the Halden project, is the second co-ordinated research programme in this area. The first step of FUMEX was to conduct a blind exercise which was carried out using 19 computer codes on 10 experimental rods irradiated in the Halden reactor. The results of this exercise are reported here and show an important improvement of modelling tools since 1984, especially due to the development of national and international parametric studies. However, shortcomings still exist and improvement in the evaluation of PCI, fuel temperature and fission gas release, especially during ramp, is not very well understood. The fuel clad gap modelling also requires further improvement. (author). 7 refs, 7 figs, 4 tabs.

Development of LWR fuel performance code FEMAXI-6

International Nuclear Information System (INIS)

Suzuki, Motoe

2006-01-01

LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

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

Science.gov (United States)

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

2013-01-01

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

Does malaria affect placental development? Evidence from in vitro models.

Directory of Open Access Journals (Sweden)

Alexandra J Umbers

Full Text Available BACKGROUND: Malaria in early pregnancy is difficult to study but has recently been associated with fetal growth restriction (FGR. The pathogenic mechanisms underlying malarial FGR are poorly characterized, but may include impaired placental development. We used in vitro methods that model migration and invasion of placental trophoblast into the uterine wall to investigate whether soluble factors released into maternal blood in malaria infection might impair placental development. Because trophoblast invasion is enhanced by a number of hormones and chemokines, and is inhibited by pro-inflammatory cytokines, many of which are dysregulated in malaria in pregnancy, we further compared concentrations of these factors in blood between malaria-infected and uninfected pregnancies. METHODOLOGY/PRINCIPAL FINDINGS: We measured trophoblast invasion, migration and viability in response to treatment with serum or plasma from two independent cohorts of Papua New Guinean women infected with Plasmodium falciparum or Plasmodium vivax in early pregnancy. Compared to uninfected women, serum and plasma from women with P. falciparum reduced trophoblast invasion (P = .06 and migration (P = .004. P. vivax infection did not alter trophoblast migration (P = .64. The P. falciparum-specific negative effect on placental development was independent of trophoblast viability, but associated with high-density infections. Serum from P. falciparum infected women tended to have lower levels of trophoblast invasion promoting hormones and factors and higher levels of invasion-inhibitory inflammatory factors. CONCLUSION/SIGNIFICANCE: We demonstrate that in vitro models of placental development can be adapted to indirectly study the impact of malaria in early pregnancy. These infections could result in impaired trophoblast invasion with reduced transformation of maternal spiral arteries due to maternal hormonal and inflammatory disturbances, which may contribute to FGR by

A Lota lota consumption: Trophic dynamics of nonnative Burbot in a valuable sport fishery

Science.gov (United States)

Klobucar, Stephen L.; Saunders, W. Carl; Budy, Phaedra

2016-01-01

Unintentional and illegal introductions of species disrupt food webs and threaten the success of managed sport fisheries. Although many populations of Burbot Lota lota are declining in the species’ native range, a nonnative population recently expanded into Flaming Gorge Reservoir (FGR), Wyoming–Utah, and threatens to disrupt predator–prey interactions within this popular sport fishery. To determine potential impacts on sport fishes, especially trophy Lake Trout Salvelinus namaycush, we assessed the relative abundance of Burbot and quantified the potential trophic or food web impacts of this population by using diet, stable isotope, and bioenergetic analyses. We did not detect a significant potential for food resource competition between Burbot and Lake Trout (Schoener’s overlap index = 0.13), but overall consumption by Burbot likely affects other sport fishes, as indicated by our analyses of trophic niche space. Diet analyses suggested that crayfish were important diet items across time (89.3% of prey by weight in autumn; 49.4% in winter) and across Burbot size-classes (small: 77.5% of prey by weight; medium: 76.6%; large: 39.7%). However, overall consumption by Burbot increases as water temperatures cool, and fish consumption by Burbot in FGR was observed to increase during winter. Specifically, large Burbot consumed more salmonids, and we estimated (bioenergetically) that up to 70% of growth occurred in late autumn and winter. Further, our population-wide consumption estimates indicated that Burbot could consume up to double the biomass of Rainbow Trout Oncorhynchus mykiss stocked annually (>1.3 × 105 kg; >1 million individuals) into FGR. Overall, we provide some of the first information regarding Burbot trophic interactions outside of the species’ native range; these findings can help to inform the management of sport fisheries if Burbot range expansion occurs elsewhere.

Spent fuel dissolution rates as a function of burnup and water chemistry

International Nuclear Information System (INIS)

Gray, W.J.

1998-06-01

To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of 129 I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and ∼ 65 MWd/kgM. (2) Oxidation of spent fuel up to the U 4 O 9+x stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of 129 I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and 129 I gap inventory for US LWR fuels

Institute for Energy Technology - Annual report 1997

International Nuclear Information System (INIS)

1998-01-01

Most of the nuclear-related work at Kjeller is based on the JEEP II research reactor, the operation of which is a prerequisite for Kjeller's activities. The work includes basic research in physics, the production of radiopharmaceuticals, the irradiation of materials for various technical applications, neutron radiography, activation analyses and silicon doping. 1997 was the first year of the 1997-1999 research period for the international OECD Halden Reactor Project. With the National Nuclear Regulatory Board (CNEN) of Brazil becoming a member late last year, the Project now include organizations from 20 countries. A membership agreement was also signed with the Institut de Protection et de Surete Nucleaire (IPSN) of France in 1997. The Project is operation by a staff of 280. In the fuel safety area, work focused chiefly on concerns arising at high burn-up in normal operation and in transient conditions. The unique Halden instrumentation has been extremely valuable for obtaining high relevance data. In respect of core materials, work has concentrated on corrosion issues, including in-core measurements of crack propagation rates in stainless steels. These data are used to estimate the expected lifetime of materials and to learn how effective measures are in improving the materials's performance. In the man-machine are the Halden Man-Machine Laboratory (HAMMLAB) is being upgraded and expanded. Modern, powerful simulators for PWR, BWR and WER reactors began to be installed during the year, and a Virtua Reality centre was set up to complement the HAMMLAB. As the infrastructure was build up, activities dwelt on human factors studies, encompassing situation awareness, the development and assessment of operator support systems, and the validation of software

Institute for Energy Technology - Annual report 1997

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-12-31

Most of the nuclear-related work at Kjeller is based on the JEEP II research reactor, the operation of which is a prerequisite for Kjeller`s activities. The work includes basic research in physics, the production of radiopharmaceuticals, the irradiation of materials for various technical applications, neutron radiography, activation analyses and silicon doping. 1997 was the first year of the 1997-1999 research period for the international OECD Halden Reactor Project. With the National Nuclear Regulatory Board (CNEN) of Brazil becoming a member late last year, the Project now include organizations from 20 countries. A membership agreement was also signed with the Institut de Protection et de Surete Nucleaire (IPSN) of France in 1997. The Project is operation by a staff of 280. In the fuel safety area, work focused chiefly on concerns arising at high burn-up in normal operation and in transient conditions. The unique Halden instrumentation has been extremely valuable for obtaining high relevance data. In respect of core materials, work has concentrated on corrosion issues, including in-core measurements of crack propagation rates in stainless steels. These data are used to estimate the expected lifetime of materials and to learn how effective measures are in improving the materials`s performance. In the man-machine are the Halden Man-Machine Laboratory (HAMMLAB) is being upgraded and expanded. Modern, powerful simulators for PWR, BWR and WER reactors began to be installed during the year, and a Virtua Reality centre was set up to complement the HAMMLAB. As the infrastructure was build up, activities dwelt on human factors studies, encompassing situation awareness, the development and assessment of operator support systems, and the validation of software.

The Multiple Roles of EG-VEGF/PROK1 in Normal and Pathological Placental Angiogenesis

Directory of Open Access Journals (Sweden)

Nadia Alfaidy

2014-01-01

Full Text Available Placentation is associated with several steps of vascular adaptations throughout pregnancy. These vascular changes occur both on the maternal and fetal sides, consisting of maternal uterine spiral arteries remodeling and placental vasculogenesis and angiogenesis, respectively. Placental angiogenesis is a pivotal process for efficient fetomaternal exchanges and placental development. This process is finely controlled throughout pregnancy, and it involves ubiquitous and pregnancy-specific angiogenic factors. In the last decade, endocrine gland derived vascular endothelial growth factor (EG-VEGF, also called prokineticin 1 (PROK1, has emerged as specific placental angiogenic factor that controls many aspects of normal and pathological placental angiogenesis such as recurrent pregnancy loss (RPL, gestational trophoblastic diseases (GTD, fetal growth restriction (FGR, and preeclampsia (PE. This review recapitulates EG-VEGF mediated-angiogenesis within the placenta and at the fetomaternal interface and proposes that its deregulation might contribute to the pathogenesis of several placental diseases including FGR and PE. More importantly this paper argues for EG-VEGF clinical relevance as a potential biomarker of the onset of pregnancy pathologies and discusses its potential usefulness for future therapeutic directions.

The multiple roles of EG-VEGF/PROK1 in normal and pathological placental angiogenesis.

Science.gov (United States)

Alfaidy, Nadia; Hoffmann, Pascale; Boufettal, Houssine; Samouh, Naima; Aboussaouira, Touria; Benharouga, Mohamed; Feige, Jean-Jacques; Brouillet, Sophie

2014-01-01

Placentation is associated with several steps of vascular adaptations throughout pregnancy. These vascular changes occur both on the maternal and fetal sides, consisting of maternal uterine spiral arteries remodeling and placental vasculogenesis and angiogenesis, respectively. Placental angiogenesis is a pivotal process for efficient fetomaternal exchanges and placental development. This process is finely controlled throughout pregnancy, and it involves ubiquitous and pregnancy-specific angiogenic factors. In the last decade, endocrine gland derived vascular endothelial growth factor (EG-VEGF), also called prokineticin 1 (PROK1), has emerged as specific placental angiogenic factor that controls many aspects of normal and pathological placental angiogenesis such as recurrent pregnancy loss (RPL), gestational trophoblastic diseases (GTD), fetal growth restriction (FGR), and preeclampsia (PE). This review recapitulates EG-VEGF mediated-angiogenesis within the placenta and at the fetomaternal interface and proposes that its deregulation might contribute to the pathogenesis of several placental diseases including FGR and PE. More importantly this paper argues for EG-VEGF clinical relevance as a potential biomarker of the onset of pregnancy pathologies and discusses its potential usefulness for future therapeutic directions.

Loschmidt echo for local perturbations: non-monotonic cross-over from the Fermi-golden-rule to the escape-rate regime

International Nuclear Information System (INIS)

Goussev, Arseni; Waltner, Daniel; Richter, Klaus; Jalabert, Rodolfo A

2008-01-01

We address the sensitivity of quantum mechanical time evolution by considering the time decay of the Loschmidt echo (LE) (or fidelity) for local perturbations of the Hamiltonian. Within a semiclassical approach, we derive analytical expressions for the LE decay for chaotic systems for the whole range from weak to strong local perturbations and identify different decay regimes which complement those known for the case of global perturbations. For weak perturbations, a Fermi-golden-rule (FGR)-type behavior is recovered. For strong perturbations, the escape-rate regime is reached, where the LE decays exponentially with a rate independent of the perturbation strength. The transition between the FGR regime and the escape-rate regime is non-monotonic, i.e. the rate of the exponential time-decay of the LE oscillates as a function of the perturbation strength. We further perform extensive quantum mechanical calculations of the LE based on numerical wave packet evolution, which strongly support our semiclassical theory. Finally, we discuss in some detail possible experimental realizations for observing the predicted behavior of the LE

Re-fabrication and Instrumentation - resume and outlook

International Nuclear Information System (INIS)

Kleeman, Hans-Joerg; Oberlaender, Barbara C.

2005-01-01

The special technique ''re-fabrication and instrumentation'' applied at the Institute for Energy Technology (IFE) makes further testing and measuring of irradiated fuel rods in the Halden Boiling Water Reactor (HBWR) possible. Machines, necessary for such operations, were designed and produced at IFE in the years 1991-92. Hot Lab-operations called ''re-fabrication'' include all modifications necessary to load an irradiated fuel rod, commercial or experimental, into the Halden reactor for further testing. ''Instrumentation'' includes all operations necessary to fit instruments into an irradiated fuel rod segment to measure for instance the temperature in the centreline of the fuel stack, pressure increase in the fuel rod and/or changes of the rod length during reactor experiments. The machines used are referred to as: ''Cutting and Grinding Unit'' for machining of the canning; ''Freezing and Drilling Unit'' for drilling of the centreline thermocouple hole in the active fuel stack;''Welding and Drying Unit'' a TIG welding machine for the circumferential welds and the seal welding; and a hydraulic ''Encapsulation Bench''. The ''Welding and Drilling Unit'' includes also a ''He-leak test chamber'' and a ''Hydraulic Press''. In addition, different types of PIE methods are used in quality assurance and documenting of the work done with the fuel rod. For example neutron radiography, visual inspection, dimension measurement, free volume measurements, equipment to measure the flow capability through the fuel stack and different function tests of the measuring devices. This paper will give an overview of the work done so far with this equipment and give information about new equipment which will be installed soon to deal with even more complicated test fuel designs. (Author)

Useful and usable alarm systems : recommended properties

International Nuclear Information System (INIS)

Veland, Oeystein; Kaarstad, Magnhild; Seim, Lars Aage; Foerdestroemmen, Nils T.

2001-01-01

This document describes the result of a study on alarm systems conducted by IFE in Halden. The study was initiated by the Norwegian Petroleum Directorate. The objective was to identify and formulate a set of important properties for useful and usable alarm systems. The study is mainly based on review of the latest international recognised guidelines and standards on alarm systems available at the time of writing, with focus on realistic solutions from research and best practice from different industries. In addition, IFE experiences gathered through specification and design of alarm systems and experimental activities in HAMMLAB and bilateral projects, have been utilized where relevant. The document presents a total of 43 recommendations divided into a number of general recommendations and more detailed recommendations on alarm generation, structuring, prioritisation, presentation and handling. (Author)

Experimental verification of stored energy calculations

International Nuclear Information System (INIS)

Hann, C.R.; Christensen, J.A.; Lanning, D.D.; Marshall, R.K.; Williford, R.E.

1975-01-01

A description is provided of irradiation tests designed to provide data needed to verify existing steady state fuel performance codes. The tests are being conducted in the Halden Reactor, and are designed to provide data pertinent to stored energy calculations over a range of linear heat ratings utilized in contemporary power reactors

PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

International Nuclear Information System (INIS)

Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

2001-01-01

A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

Annual report 1986 - IFE

International Nuclear Information System (INIS)

1987-01-01

At present Norwegian nuclear energy research is centered around the international OECD Halden Reactor Project which has participants from 40 organisations in 10 countries, including the Nordic countries and most major nuclear power countries within OECD. The research programme is concentrated on nuclear fuel and safety technology, and on computerbased methods for operation and supervision of power reactors. The paramount objective of fuel research at Halden is to produce experimental data that will increase insight into the fundamental factors affecting the reliability of the fuel under various operational conditions, and to obtain data for developing and qualifying analytical/empirical calculation models of fuel behaviour. Joint programme activities are increasingly being directed towards defining long-term effects that may influence the fuel's operational reliability. To follow up these progress trends, a special loop was installed in the reactor in 1986 for studying conditions affecting corrosion of fuel rods. Of prime significance in the development work is the advanced experimental control room which is connected to a full-scale simulator of a nuclear power plant of the PWR type. In 1986 the experimental control room was used for comparative studies of alarm systems based on conventional alarm displays and an advanced system developed by the Halden Project which are based on a colour screen display. The latter system eliminates alarms that are irrelevant to the current plant status, so reducing the volume of information reaching the operator. In 1986 work was also done on a major computerized system designed to assist operators under irregular operating conditions. The Project developed a module for early detection of irregularities, an expert diagnosis system to find causes, and a module whereby the operator is led through the procedures that must be followed in order to bring the process back to an acceptable state

Institutt for Energiteknikk - Annual report 1992

International Nuclear Information System (INIS)

1993-01-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. The main nuclear program is centered on the Halden Reactor Project. In 1958, the first Halden Reactor Project Agreement was signed by organisations representing 12 European countries. The project's membership now includes several associated parties in addition to 13 organisations from its 12 NEA members plus one NEA non-member, the Czech and Slovak Federal Republic. The objectives have evolved from being simply a demonstration of the operation of a boiling heavy-water reactor to becoming a substantial research and development programme covering the domains of human-machine interaction, fuel behaviour, materials testing, water chemistry, and instrumentation. In 1992, significant progress was achieved in all of the areas addressed by the project, including the re-instrumentation of irradiated fuel rods, fission gas release, irradiation-assisted stress corrosion cracking, a conceptual design of advanced cockpit-type control rooms, analysis of human behaviour, and information processing and presentation. More than 33 years after its reactor's first criticality, the Halden Project continues to provide the scientific community with a wealth of information for the safe and efficient production of nuclear power. The current Project Agreement covers the period from 1990 to 1993, and a further three-year extension to 1996 is being discussed among the members. At Kjeller irradiation services are based on the JEEP II reactor and the gamma irradiation facility. Neutron irradiation in the reactor produces the radioactive raw materials for the manufacture of radiopharmaceuticals and radiochemicals. Neutron irradiation is also used for the accurate control of conductivity in superpure silicon crystals. The main purpose of IFE's basic research in physics is to utilize neutron beams from the JEEP II reactor for fundamental studies of the physical characteristics of solids and complex liquids

A fuel thermal conductivity correlation based on the latest experimental results

International Nuclear Information System (INIS)

Sontheimer, F.; Landskron, H.; Billaux, M.R.

2000-01-01

A new fuel thermal conductivity (ftc) correlation for UO 2 and (U,Gd)O 2 is presented, which is based on the relaxation-time theory of Klemens. The correlation is chosen because of its validity in a wide range of defect concentrations as for instance encountered in fuel with a wide range of burnup and gadolinia additions, as has been shown by Ishimoto. The phonon term of the new correlation has the form 1/x·arctan(x) , where x is a measure of the defect concentration introduced by burnup and gadolinia additions. For low defect concentrations, this term is identical with the classical form for the phonon term 1/(A+B.T). At high defect concentrations, however, when phonon-point defect scattering starts dominating over phonon-phonon scattering, the new correlation deviates from the classical formulation and has a distinctly weaker dependence on temperature and defect concentration than the classical form. The new arctan correlation in combination with an appropriate electronic ftc term is fitted to the Halden data base of fuel centre-line temperature measurements (represented by the ''Halden ftc correlation recommendation''). Agreement is very good up to a burnup of about 60 MWd/kgU; beyond, the arctan form has a saturating burnup degradation. The new arctan correlation in combination with an appropriate electronic ftc term is also shown to describe very well our latest ftc measurements on unirradiated gadolinia fuel up to 9% gadolinia content. Application to Halden measurements up to very high burnup is successful, when combined with the so-called ''rim-effect'', which counteracts the saturation tendency of the new correlation at high burnup. Latest laser thermal diffusivity measurements on irradiated gadolinia fuel in the frame of the NFIR program, although not yet open for literature and not discussed in the paper, indicate very good agreement with the new arctan correlation. (author)

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2012-01-01

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

LVDT Development for High Temperature Irradiation Test and Application

Energy Technology Data Exchange (ETDEWEB)

Lee, Chul Yong; Ban, Chae Min; Choo, Kee Nam; Jun, Byung Hyuk [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The LVDT (Linear Variable Differential Transformer) is used to measure the elongation and pressure of a nuclear fuel rod, or the creep and fatigue of the material during a reactor irradiation test. This device must be a radiation-resistant LVDT for use in a research reactor. Norway Halden has LVDTs for an irradiation test by the own development and commercialized. But Halden's LVDTs have limited the temperature of the use until to 350 .deg. C. So, KAERI has been developing a new LVDT for high temperature irradiation test. This paper describes the design of a LVDT, the fabrication process of a LVDT, and the result of the performance test. The designed LVDT uses thermocouple cable for coil wire material and one MI cable as signal cable. This LVDT for a high temperature irradiation test can be used until a maximum of 900 .deg. C. Welding is a very important factor for the fabrication of an LVDT. We are using a 150W fiber laser welding system that consists of a welding head, monitoring vision system and rotary index.

Model checking as an aid to procedure design

International Nuclear Information System (INIS)

Zhang, Wenhu

2001-01-01

The OECD Halden Reactor Project has been actively working on computer assisted operating procedures for many years. The objective of the research has been to provide computerised assistance for procedure design, verification and validation, implementation and maintenance. For the verification purpose, the application of formal methods has been considered in several reports. The recent formal verification activity conducted at the Halden Project is based on using model checking to the verification of procedures. This report presents verification approaches based on different model checking techniques and tools for the formalization and verification of operating procedures. Possible problems and relative merits of the different approaches are discussed. A case study of one of the approaches is presented to show the practical application of formal verification. Application of formal verification in the traditional procedure design process can reduce the human resources involved in reviews and simulations, and hence reduce the cost of verification and validation. A discussion of the integration of the formal verification with the traditional procedure design process is given at the end of this report. (Author)

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

International Nuclear Information System (INIS)

2002-01-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels

Beginning-of-Life Data Report for the Instrumented Fuel Assembly (IFA)-527

Energy Technology Data Exchange (ETDEWEB)

Lanning, D. D.

1981-09-01

This report presents beginning-of-life (BOL) data from the first four months of operation of the six-rod instrumented fuel assembly (IFA)-527 in the Halden Boiling Water Reactor (HBWR), Halden, Norway. This assembly is the last in a series of U.S. Nuclear Regulatory Commission (NRC)-sponsored tests to verify steady-state fuel performance computer codes. IFA-527 contains five identical rods with high-density stable fuel pellets and 0.23-mm diametral gaps and one rod with similar fuel pellets but with a 0.06-mm diametral gap. All six rods were xenon-filled to provide simulation of the effects of fission gas and to enhance the observable effects of fuel cracking and relocation on fuel temperatures. The assembly operated successfully from July 1, 1980, to August 15, 1980; and then the reactor was shut down until September 10, 1980. Sometime during the shutdown, four of the six rods suffered pressure boundary failure. The decision was made to restart the reactor to collect operating data with failed rods. This report presents both pre- and postfailure data for IFA-527.

LOCA testing of high burnup PWR fuel in the HBWR. Additional PIE on the cladding of the segment 650-5

Energy Technology Data Exchange (ETDEWEB)

Oberlaender, B.C.; Espeland, M.; Jenssen, H.K.

2008-07-01

IFA-650.5, a test with pre-irradiated fuel in the Halden Project LOCA test series, was conducted on October 23rd, 2006. The fuel rod had been used in a commercial PWR and had a high burnup, 83 MWd/kgU. Experimental arrangements of the fifth test were similar to the preceding LOCA tests. The peak cladding temperature (PCT) level was higher than in the third and fourth tests, 1050 C. A peak temperature close to the target was achieved and cladding burst occurred at approx. 750 C. Within the joint programme framework of the Halden Project PIE was done, consisting of gamma scanning, visual inspection, neutron-radiography, hydrogen analysis and metallography / ceramography. An additional extensive PIE including metallography, hydrogen analysis, and hardness measurements of cross-sections at seven axial elevations was done. It was completed to study the high burnup and LOCA induced effects on the Zr-4 cladding, namely the migration of oxygen into the cladding from the inside surface, the cladding distension, and the burst (author)(tk)

Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies

International Nuclear Information System (INIS)

2001-01-01

In co-operation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on ''Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies'' in the period August 27-August 31, 2001. The Summer School was intended for scientists, engineers and technicians working for nuclear installations, engineering companies, industry and members of universities and research institutes, who wanted to broaden their nuclear background by getting acquainted with Man-Technology-Organisation-related subjects and issues. The Summer School should also serve to transfer knowledge to the ''young generation'' in the nuclear field. The following presentations were given: (1) Overview of the Nuclear Community and Current issues, (2) The Elements of Safety Culture; Evaluation of Events, (3) Quality Management (QM), (4) Probabilistic Risk Assessment (PSA), (5) Human Behaviour from the Viewpoint of Industrial Psychology, (6) Technical tour of the Halden Project Experimental Facilities, (7) Human Factors in Control Room Design, (8) Computerised Operator Support Systems (COSSs) and (9) Artificial Intelligence; a new Approach. Most of the contributions are overhead figures from spoken lectures

Validation of radiation dose estimations in VRdose: comparing estimated radiation doses with observed radiation doses

International Nuclear Information System (INIS)

Nystad, Espen; Sebok, Angelia; Meyer, Geir

2004-04-01

The Halden Virtual Reality Centre has developed work-planning software that predicts the radiation exposure of workers in contaminated areas. To validate the accuracy of the predicted radiation dosages, it is necessary to compare predicted doses to actual dosages. During an experimental study conducted at the Halden Boiling Water Reactor (HBWR) hall, the radiation exposure was measured for all participants throughout the test session, ref. HWR-681 [3]. Data from this experimental study have also been used to model tasks in the work-planning software and gather data for predicted radiation exposure. Two different methods were used to predict radiation dosages; one method used all radiation data from all the floor levels in the HBWR (all-data method). The other used only data from the floor level where the task was conducted (isolated data method). The study showed that the all-data method gave predictions that were on average 2.3 times higher than the actual radiation dosages. The isolated-data method gave predictions on average 0.9 times the actual dosages. (Author)

Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

In co-operation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on ''Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies'' in the period August 27-August 31, 2001. The Summer School was intended for scientists, engineers and technicians working for nuclear installations, engineering companies, industry and members of universities and research institutes, who wanted to broaden their nuclear background by getting acquainted with Man-Technology-Organisation-related subjects and issues. The Summer School should also serve to transfer knowledge to the ''young generation'' in the nuclear field. The following presentations were given: (1) Overview of the Nuclear Community and Current issues, (2) The Elements of Safety Culture; Evaluation of Events, (3) Quality Management (QM), (4) Probabilistic Risk Assessment (PSA), (5) Human Behaviour from the Viewpoint of Industrial Psychology, (6) Technical tour of the Halden Project Experimental Facilities, (7) Human Factors in Control Room Design, (8) Computerised Operator Support Systems (COSSs) and (9) Artificial Intelligence; a new Approach. Most of the contributions are overhead figures from spoken lectures.

NOMAGE4 activities 2011. Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activities in 2011

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (Norway))

2012-01-15

A network for materials issues has been initiated in 2009 within the Nordic countries. The original objectives of the Generation IV Nordic Nuclear Materials Forum (NOMAGE4) were to form the basis of a sustainable forum for Gen-IV issues, especially focusing on fuels, cladding, structural materials and coolant interaction. Over the last years, other issues such as reactor physics, thermal hydraulics, safety and waste have gained in importance (within the network) and therefore the scope of the forum has been enlarged and a more appropriate and more general name, NORDIC-GEN4, has been chosen for the forum. Further, the interaction with non-Nordic countries (such as The Netherlands (JRC, NRG) and Czech Republic (CVR)) will be increased. Within the NOMAGE4 project, a seminar was organized by IFE-Halden during 31 October - 1 November 2011. The seminar attracted 65 participants from 12 countries. The seminar provided a forum for exchange of information, discussion on future research reactor needs and networking of experts on Generation IV reactor concepts. The participants could also visit the Halden reactor site and the workshop. (Author)

Institutt for Atomenergi

International Nuclear Information System (INIS)

Anon.

1977-01-01

The research programme of Institutt for Atomenergi is described giving brief descriptions of some of the work done in the fields of reactor safety, the operation and inspection of the Halden reactor, fuel cycle calculations, fuel element performance testing, isotope production and applications, the production of high purity gadolinium oxide, and fundamental research in crystallography. (JIW)Ψ

National report Netherlands KEMA

International Nuclear Information System (INIS)

Loriaux, E.F.

1995-01-01

COPMA is a product of the OECD Halden Reactor Program (HRP). It is a COmputerized Procedures MAnual. The product is in an evaluation phase. Through KEMA's associated membership this product was evaluated by representatives of the two Dutch Nuclear Power plants, EPZ PWR plant in Borssele and GKN BWR plant Dodewaard. This presentation is a summary of the report

Energy diffusion in strongly driven quantum chaotic systems: the role of correlations of the matrix elements

International Nuclear Information System (INIS)

Elyutin, P V; Rubtsov, A N

2008-01-01

The energy evolution of a quantum chaotic system under the perturbation that harmonically depends on time is studied for the case of large perturbation, in which the rate of transition calculated from the Fermi golden rule (FGR) is about or exceeds the frequency of perturbation. For this case, the models of the Hamiltonian with random non-correlated matrix elements demonstrate that the energy evolution retains its diffusive character, but the rate of diffusion increases slower than the square of the magnitude of perturbation, thus destroying the quantum-classical correspondence for the energy diffusion and the energy absorption in the classical limit ℎ → 0. The numerical calculation carried out for a model built from the first principles (the quantum analog of the Pullen-Edmonds oscillator) demonstrates that the evolving energy distribution, apart from the diffusive component, contains a ballistic one with the energy dispersion that is proportional to the square of time. This component originates from the chains of matrix elements with correlated signs and vanishes if the signs of matrix elements are randomized. The presence of the ballistic component formally extends the applicability of the FGR to the non-perturbative domain and restores the quantum-classical correspondence

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

International Nuclear Information System (INIS)

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

2003-05-01

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

Annual report 1984

International Nuclear Information System (INIS)

Nuclear power technology, isotope technology and basic research in physics comprises the nuclear activities of Institute for Energy Technology (IFE). In 1984 the Institute entered into an agreement with the present foreign signatories (comprising more than 40 organisations from 10 participating countries) for the continuation of the OECD Halden Project for the three year period from 1985 to 1987. IFE's share of the total budget is appr. 25%. The project's fuel investigations have in 1984 particularly emphasised the characterisation of long term effects with regard to efficiency and operational safety, and the determination of reliability and durability in the case of accident with the loss of cooling water. During 1984 the new experimental reactor control room of the Halden Project was put in operation. The screen based control room, connected to a detailed simulator of a PWR, is intended for experiments in the field of man-machine communication, information presentation, testing of specific systems and training of operators. A testing of alarm systems carried out during 1984 shows that the filtering of alarms is an effective means to ensure that the operator concentrates on the most relevant information during disruptions in the plant operations, and that the use of mimic diagrams facilitates the indentification of alarm situations. Apart from isotope production, the efforts of the isotope laboratories have been concentrated on the development of two completely new radiopharmaceuticals, with the aid of which a blood sample can be marked with a radioactive tracer for the location of abcesses, states of inflammation or blood clots. Research in basic physics has mainly been based on the use of neutron irradiations from the JEEP II reactor for the investigation of the structure and dynamics of solid state materials

Annual report 1990

International Nuclear Information System (INIS)

1991-01-01

Work at Institutt for energiteknikk (IFE) comprises both nuclear and non-nuclear activities. Nuclear power research at the Institute is performed within the international OECD Halden Reactor Project. More than 40 organizations in eight European countries, plus USA and Japan, participated in the 1988 to 1990 phase of the project. The next agreed phase, 1991 to 1993, will see the list of participants expanded to include two new OECD countries, Spain and Switzerland. Key factors under investigation in the fuel and materials technology area include the operational reliablility of reactor fuel and changes in material properties under varying conditions of radiation and water chemistry. Progress has been satisfactory on the fuel and materials test program, and about 40 different fuel element designs have been tested. The information technology activities centre on an advanced control room coupled to a full scale simulator of a nuclear power station of the pressurized water type (PWT). The experiments, aimed at developing new informations systems for energy and process industries, are carried out by process operators from the Halden reactor. During 1990, an extensive program has been carried out to test a computer-based operation procedure system. The results show that this system, in certain operational situations, provides the operator with better information support than conventional handbooks of written procedures. A number of areas in which the system could be improved were also identified, and the development of a new ''second generation'' system was started in 1990. The new system will be implemented on UNIX-based work stations, using so-called open system standards in order to facilitate transfer of the system to different users. Testing has also been started on an operator support system for diagnosis on the primary causes of plant disturbances. 22 figs

Real-time 3D radiation risk assessment supporting simulation of work in nuclear environments

International Nuclear Information System (INIS)

Szoke, I; Louka, M N; Bryntesen, T R; Bratteli, J; Edvardsen, S T; RøEitrheim, K K; Bodor, K

2014-01-01

This paper describes the latest developments at the Institute for Energy Technology (IFE) in Norway, in the field of real-time 3D (three-dimensional) radiation risk assessment for the support of work simulation in nuclear environments. 3D computer simulation can greatly facilitate efficient work planning, briefing, and training of workers. It can also support communication within and between work teams, and with advisors, regulators, the media and public, at all the stages of a nuclear installation’s lifecycle. Furthermore, it is also a beneficial tool for reviewing current work practices in order to identify possible gaps in procedures, as well as to support the updating of international recommendations, dissemination of experience, and education of the current and future generation of workers. IFE has been involved in research and development into the application of 3D computer simulation and virtual reality (VR) technology to support work in radiological environments in the nuclear sector since the mid 1990s. During this process, two significant software tools have been developed, the VRdose system and the Halden Planner, and a number of publications have been produced to contribute to improving the safety culture in the nuclear industry. This paper describes the radiation risk assessment techniques applied in earlier versions of the VRdose system and the Halden Planner, for visualising radiation fields and calculating dose, and presents new developments towards implementing a flexible and up-to-date dosimetric package in these 3D software tools, based on new developments in the field of radiation protection. The latest versions of these 3D tools are capable of more accurate risk estimation, permit more flexibility via a range of user choices, and are applicable to a wider range of irradiation situations than their predecessors. (paper)

Real-time 3D radiation risk assessment supporting simulation of work in nuclear environments.

Science.gov (United States)

Szőke, I; Louka, M N; Bryntesen, T R; Bratteli, J; Edvardsen, S T; RøEitrheim, K K; Bodor, K

2014-06-01

This paper describes the latest developments at the Institute for Energy Technology (IFE) in Norway, in the field of real-time 3D (three-dimensional) radiation risk assessment for the support of work simulation in nuclear environments. 3D computer simulation can greatly facilitate efficient work planning, briefing, and training of workers. It can also support communication within and between work teams, and with advisors, regulators, the media and public, at all the stages of a nuclear installation's lifecycle. Furthermore, it is also a beneficial tool for reviewing current work practices in order to identify possible gaps in procedures, as well as to support the updating of international recommendations, dissemination of experience, and education of the current and future generation of workers.IFE has been involved in research and development into the application of 3D computer simulation and virtual reality (VR) technology to support work in radiological environments in the nuclear sector since the mid 1990s. During this process, two significant software tools have been developed, the VRdose system and the Halden Planner, and a number of publications have been produced to contribute to improving the safety culture in the nuclear industry.This paper describes the radiation risk assessment techniques applied in earlier versions of the VRdose system and the Halden Planner, for visualising radiation fields and calculating dose, and presents new developments towards implementing a flexible and up-to-date dosimetric package in these 3D software tools, based on new developments in the field of radiation protection. The latest versions of these 3D tools are capable of more accurate risk estimation, permit more flexibility via a range of user choices, and are applicable to a wider range of irradiation situations than their predecessors.

Proceedings of the Workshop on Current and Emerging methods for Optimising Safety and Efficiency in Nuclear Decommissioning

International Nuclear Information System (INIS)

2017-02-01

The workshop was organised by the Institute for Energy Technology (IFE) on behalf of the OECD Halden Reactor Project (OECD-HRP) and in collaboration with the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency (NEA). The workshop brought together more than 100 people (operators, regulators, scientists, consultants, and contractors) from 25 countries. Program: Day 1 - Successful application of R and D in decommissioning and future needs (Welcome and Opening Speeches; Session 1: Workshop Introductory Presentations; Session 2: Experience from starting, on-going and completed decommissioning projects). Day 2 - R and D and application of advanced technologies for decommissioning (Session 3: New technologies for decommissioning; Session 4: Advanced information technologies for decommissioning). Day 3 - Improving decommissioning management on project, national and international level (Session 5: Challenges and methods for improving decommissioning; Session 6: Workshop closing)

Vanadium Beta Emission Detectors for Reactor In-Core Neutron Monitoring

Energy Technology Data Exchange (ETDEWEB)

Andersson, I Oe; Soederlund, B

1969-06-15

In-core flux measurements are becoming increasingly important in both power reactors and test reactors. In particular power distribution measurements in large power reactors have to be performed with a great number of neutron detectors capable of withstanding high integrated flux values. This report presents a summary of the development and application of a new type of nuclear radiation sensor, a beta emission detector, for measurements at high neutron flux levels. The work has been carried out at the Section for Instrumentation and has been the basis for a type of neutron detector employed in the Marviken in-core system as well as for other types. The report describes the design and principle of operation, experiments and tests. Also included are the results and comments from a long-term irradiation of some detectors in the Halden reactor.

Compliance characteristics of cracked UO2 pellets

International Nuclear Information System (INIS)

Williford, R.E.; Mohr, C.L.; Lanning, D.D.

1981-01-01

The thermally induced cracking of UO 2 fuel pellets causes simultaneous reductions of the bulk (extrinsic) fuel thermal conductivity and elastic moduli to values significantly less than those for solid pellets. The magnitude of these bulk properly reductions was found to be primarily dependent on the amount of crack area in the transverse plane of the fuel. The model described herein uses a simple description of the crack geometry to couple the fuel rod thermal and mechanical behaviors by relating in-reactor data to Hooke's Law and a crack compliance model. Data from the NRC/PNL Halden experiment IFA-432 show that for a typical helium-filled BWR-design rod at 30 kW/m, the effective thermal conductivity and elastic moduli of the cracked fuel are 4/5 and 1/40 of that for solid pellets, respectively

Melatonin rescues cardiovascular dysfunction during hypoxic development in the chick embryo

OpenAIRE

Itani, Nozomi; Skeffington, Katie L.; Beck, Christian; Niu, Youguo; Giussani, Dino A.

2015-01-01

Abstract There is a search for rescue therapy against fetal origins of cardiovascular disease in pregnancy complicated by chronic fetal hypoxia, particularly following clinical diagnosis of fetal growth restriction (FGR). Melatonin protects the placenta in adverse pregnancy; however, whether melatonin protects the fetal heart and vasculature in hypoxic pregnancy independent of effects on the placenta is unknown. Whether melatonin can rescue fetal cardiovascular dysfunction when treatment comm...

Exploring Chondrule and CAI Rims Using Micro- and Nano-Scale Petrological and Compositional Analysis

Science.gov (United States)

Cartwright, J. A.; Perez-Huerta, A.; Leitner, J.; Vollmer, C.

2017-12-01

As the major components within chondrites, chondrules (mm-sized droplets of quenched silicate melt) and calcium-aluminum-rich inclusions (CAI, refractory) represent the most abundant and the earliest materials that solidified from the solar nebula. However, the exact formation mechanisms of these clasts, and whether these processes are related, remains unconstrained, despite extensive petrological and compositional study. By taking advantage of recent advances in nano-scale tomographical techniques, we have undertaken a combined micro- and nano-scale study of CAI and chondrule rim morphologies, to investigate their formation mechanisms. The target lithologies for this research are Wark-Lovering rims (WLR), and fine-grained rims (FGR) around CAIs and chondrules respectively, present within many chondrites. The FGRs, which are up to 100 µm thick, are of particular interest as recent studies have identified presolar grains within them. These grains predate the formation of our Solar System, suggesting FGR formation under nebular conditions. By contrast, WLRs are 10-20 µm thick, made of different compositional layers, and likely formed by flash-heating shortly after CAI formation, thus recording nebular conditions. A detailed multi-scale study of these respective rims will enable us to better understand their formation histories and determine the potential for commonality between these two phases, despite reports of an observed formation age difference of up to 2-3 Myr. We are using a combination of complimentary techniques on our selected target areas: 1) Micro-scale characterization using standard microscopic and compositional techniques (SEM-EBSD, EMPA); 2) Nano-scale characterization of structures using transmission electron microscopy (TEM) and elemental, isotopic and tomographic analysis with NanoSIMS and atom probe tomography (APT). Preliminary nano-scale APT analysis of FGR morphologies within the Allende carbonaceous chondrite has successfully discerned

Placental oxidative stress and decreased global DNA methylation are corrected by copper in the Cohen diabetic rat

Energy Technology Data Exchange (ETDEWEB)

Ergaz, Zivanit, E-mail: zivanit@hadassah.org.il [Hebrew University Hadassah Medical School, Jerusalem (Israel); Guillemin, Claire [Department of Pharmacology and Therapeutics, McGill University, Montreal (Canada); Neeman-azulay, Meytal; Weinstein-Fudim, Liza [Hebrew University Hadassah Medical School, Jerusalem (Israel); Stodgell, Christopher J.; Miller, Richard K. [Department of Obstetrics and Gynecology, University of Rochester, Rochester (United States); Szyf, Moshe [Department of Pharmacology and Therapeutics, McGill University, Montreal (Canada); Ornoy, Asher [Hebrew University Hadassah Medical School, Jerusalem (Israel)

2014-05-01

Fetal Growth Restriction (FGR) is a leading cause for long term morbidity. The Cohen diabetic sensitive rats (CDs), originating from Wistar, develop overt diabetes when fed high sucrose low copper diet (HSD) while the original outbred Sabra strain do not. HSD induced FGR and fetal oxidative stress, more prominent in the CDs, that was alleviated more effectively by copper than by the anti-oxidant vitamins C and E. Our aim was to evaluate the impact of copper or the anti-oxidant Tempol on placental size, protein content, oxidative stress, apoptosis and total DNA methylation. Animals were mated following one month of HSD or regular chow diet and supplemented throughout pregnancy with either 0, 1 or 2 ppm of copper sulfate or Tempol in their drinking water. Placental weight on the 21st day of pregnancy decreased in dams fed HSD and improved upon copper supplementation. Placental/fetal weight ratio increased among the CDs. Protein content decreased in Sabra but increased in CDs fed HSD. Oxidative stress biochemical markers improved upon copper supplementation; immunohistochemistry for oxidative stress markers was similar between strains and diets. Caspase 3 was positive in more placentae of dams fed HSD than those fed RD. Placental global DNA methylation was decreased only among the CDs dams fed HSD. We conclude that FGR in this model is associated with smaller placentae, reduced DNA placental methylation, and increased oxidative stress that normalized with copper supplementation. DNA hypomethylation makes our model a unique method for investigating genes associated with growth, oxidative stress, hypoxia and copper. - Highlights: • Sensitive Cohen diabetic rats (CDs) had small placentae and growth restricted fetuses. • CDs dams fed high sucrose low copper diet had placental global DNA hypomethylation. • Caspase 3 was positive in more placentae of dams fed HSD than those fed RD. • Oxidative stress parameters improved by Tempol and resolved by copper

Spatial distribution of soda straws growth rates of the Coufin Cave (Vercors, France

Directory of Open Access Journals (Sweden)

Perrette Yves

2010-07-01

Full Text Available The Choranche Cave system (Vercors, France is an excellent locality for measuring the growth rates of large numbers soda straws. This is especially the case for the Coufin Cave, as enlargement of the cave entrance in 1875 led to a change in stalactite color from brown to white, thus providing a reliable chronomarker. The date of this brown-to-white calcite transition has been confirmed by lamina counting. We measured and georeferenced the growth-lengths of 306 soda straws in a 1m2 area of the roof of the Coufin Cave entrance chamber. Because of the very slow and sometimes inexistent water feeding of those stalactites, hydrochemistry analysis were not achieved and drop rate effect on growth were neglected; this study is based on a geomorphological and geostatistical work. By measuring a large number of soda straws in a very small area for which most of the parameters affecting stalactite growth could be considered uniform, and because flow rates are very slow (frequencies are always superior to 1 drop per half hour, we could ascribe differences in growth rates to variations in the global increase of water flow through the unsaturated matrix. Statistical and geostatistical analyses of the measurements showed that this set of similarly shaped stalactites actually consisted of three Gaussian populations with different mean growth rates: fast growth rate (FGR- mean of 0.92 mm.y-1, medium growth rate (MGR- mean of 0.47 mm.y-1 and low growth rate (LGR- 0.09 mm.y-1. Plotting the lengths and spatial distribution of the 20 longest FGR soda straws revealed that there is a rough pattern to the water flow through the cave roof. Even if no direction is statisticaly different from others, the observed directional pattern is consistent with local and regional tectonic observations. Plots of the spatial distribution of the soda straws show that FGR soda straws follow lines of regional geological stress, whereas MGR and LGR soda straws are more dispersed.

Placental oxidative stress and decreased global DNA methylation are corrected by copper in the Cohen diabetic rat

International Nuclear Information System (INIS)

Ergaz, Zivanit; Guillemin, Claire; Neeman-azulay, Meytal; Weinstein-Fudim, Liza; Stodgell, Christopher J.; Miller, Richard K.; Szyf, Moshe; Ornoy, Asher

2014-01-01

Fetal Growth Restriction (FGR) is a leading cause for long term morbidity. The Cohen diabetic sensitive rats (CDs), originating from Wistar, develop overt diabetes when fed high sucrose low copper diet (HSD) while the original outbred Sabra strain do not. HSD induced FGR and fetal oxidative stress, more prominent in the CDs, that was alleviated more effectively by copper than by the anti-oxidant vitamins C and E. Our aim was to evaluate the impact of copper or the anti-oxidant Tempol on placental size, protein content, oxidative stress, apoptosis and total DNA methylation. Animals were mated following one month of HSD or regular chow diet and supplemented throughout pregnancy with either 0, 1 or 2 ppm of copper sulfate or Tempol in their drinking water. Placental weight on the 21st day of pregnancy decreased in dams fed HSD and improved upon copper supplementation. Placental/fetal weight ratio increased among the CDs. Protein content decreased in Sabra but increased in CDs fed HSD. Oxidative stress biochemical markers improved upon copper supplementation; immunohistochemistry for oxidative stress markers was similar between strains and diets. Caspase 3 was positive in more placentae of dams fed HSD than those fed RD. Placental global DNA methylation was decreased only among the CDs dams fed HSD. We conclude that FGR in this model is associated with smaller placentae, reduced DNA placental methylation, and increased oxidative stress that normalized with copper supplementation. DNA hypomethylation makes our model a unique method for investigating genes associated with growth, oxidative stress, hypoxia and copper. - Highlights: • Sensitive Cohen diabetic rats (CDs) had small placentae and growth restricted fetuses. • CDs dams fed high sucrose low copper diet had placental global DNA hypomethylation. • Caspase 3 was positive in more placentae of dams fed HSD than those fed RD. • Oxidative stress parameters improved by Tempol and resolved by copper

Supervision of Norwegian nuclear pants 2009-2011; Tilsyn med norske atomanlegg 2009-2011

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-01

Institute for Energy Technology (IFE), is responsible for Norway's nuclear power plants in Halden, at Kjeller and Himdalen. NRPA causes continuous supervision of all aspects of the IFE's activities, among other things, nuclear safety, emergency preparedness and emissions. In the period 2009-2011 it has not been any serious incidents at IFE's nuclear power plants.(Author)

STEM - software test and evaluation methods. A study of failure dependency in diverse software

International Nuclear Information System (INIS)

Bishop, P.G.; Pullen, F.D.

1989-02-01

STEM is a collaborative software reliability project undertaken in partnership with Halden Reactor Project, UKAEA, and the Finnish Technical Research Centre. The objective of STEM is to evaluate a number of fault detection and fault estimation methods which can be applied to high integrity software. This Report presents a study of the observed failure dependencies between faults in diversely produced software. (author)

Status of the USNRC/PNL Halden test IFA-432

International Nuclear Information System (INIS)

Bradley, E.R.; Cunningham, M.E.; Lanning, D.D.; Williford, R.E.

1980-01-01

The NRC/PNL fuel assembly IFA-432 began irradiation in December 1975 and as of January 1980 had reached an assembly average burnup of 2120 GJ/kgU (24.5 GWd/MTM). The rods in this assembly are heavily instrumented and this has allowed a fairly thorough analysis of fuel behavior. The emphasis of the analysis has been to evaluate the observed changes in thermal/mechanical behavior. These changes are related to estimates of fuel density changes, fission gas release, and fuel cracking and relocation. The following conclusions have been reached concerning the irradiation of IFA-432. Based on analysis of steady-state and transient thermal data, the fuel in all rods is extensively cracked and relocated. Significant reductions in the effective fuel thermal conductivity have been deduced from the analysis of the transient thermal data

Demonstrate use of capillary electrophoresis low level transient of anions

International Nuclear Information System (INIS)

Moum, Kari-Lye; Solheim, Torill; McElrath, Joel; Frattini, Paul

2012-09-01

Capillary Electrophoresis (CE) is a well-known analytical method capable of rapid detection of very low concentration of cations and anionic species such as chloride, sulfate and nitrate. These anions are of crucial importance in reducing the potential of stainless steel components to undergo stress corrosion cracking. Currently, Nuclear Power Plants (NPPs) use Ion Chromatography (IC) as the analytical technique to achieve the required detection levels of ionic species. At the Halden Reactor Project (HRP) IC was replaced by CE in 1996, and since then HRP has gained nearly 20 years of operational experience. During the last 15 years, EPRI has done research on the CE technique and has achieved extensive experience in this area. EPRI has demonstrated detection levels at ppt and sub-ppb levels. This paper presents the ability of the CE technique to follow low level transients of anions in Boiling Water Reactor (BWR) coolant. A transient caused by approx. 10 ppb chloride and sulfate was simulated in an experimental circuit simulating BWR conditions. A series of grab samples were taken and analysed using HRPs CE (Agilent G1600). (authors)

Computerized accident management support system: development for severe accident management

International Nuclear Information System (INIS)

Garcia, V.; Saiz, J.; Gomez, C.

1998-01-01

The activities involved in the international Halden Reactor Project (HRP), sponsored by the OECD, include the development of a Computerized Accident Management Support System (CAMS). The system was initially designed for its operation under normal conditions, operational transients and non severe accidents. Its purpose is to detect the plant status, analyzing the future evolution of the sequence (initially using the APROS simulation code) and the possible recovery and mitigation actions in case of an accident occurs. In order to widen the scope of CAMS to severe accident management issues, the integration of the MAAP code in the system has been proposed, as the contribution of the Spanish Electrical Sector to the project (with the coordination of DTN). To include this new capacity in CAMS is necessary to modify the system structure, including two new modules (Diagnosis and Adjustment). These modules are being developed currently for Pressurized Water Reactors and Boiling Water REactors, by the engineering of UNION FENOSA and IBERDROLA companies (respectively). This motion presents the characteristics of the new structure of the CAMS, as well as the general characteristics of the modules, developed by these companies in the framework of the Halden Reactor Project. (Author)

A methodology and status of technology for fault diagnosis

International Nuclear Information System (INIS)

Kim, Jung Taek; Ham, Chang Shik; Kwon, Kee Choon; Lee, Dong Young; Hwang, In Koo; Song, Soon Ja; Park, Joo Hweon

1998-05-01

Since the 1980's, a nuclear industry has been attempting to apply an artificial intelligent system into MMIS. Such attempts have, especially, been led by U.S.A., Japan and Halden, which were more active for studying an artificial intelligent system. a diagnostic system is being developed such a small system that is the more frequent faults or directly effects fault into a operation and a safety of plants. Such a small diagnostic system gives a diagnostic information into the alarm processing systems or the plant information monitoring systems as integrated with these large systems. There are two major methods of diagnosis of faults. The first method is to make a misbehavior on components or processes into knowledge base and the other is to make a misbehavior on components or processes into processing model. The latter has the advantage of the former. There are OASYS and alarm processing system in ADIOS as the typical diagnostic systems on knowledge base. There are MOAS-II, Halden's diagnostic systems and diagnostic model in ADIOS as the typical diagnostic systems on model base. (author). 32 refs., 18 tabs., 28 figs

Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

In co-operation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on ''Safe Operation of Nuclear Power Plants: Impacts of Human and Organisational Factors and Emerging Technologies'' in the period August 27-August 31, 2001. The Summer School was intended for scientists, engineers and technicians working for nuclear installations, engineering companies, industry and members of universities and research institutes, who wanted to broaden their nuclear background by getting acquainted with Man-Technology-Organisation-related subjects and issues. The Summer School should also serve to transfer knowledge to the ''young generation'' in the nuclear field. The following presentations were given: (1) Overview of the Nuclear Community and Current issues, (2) The Elements of Safety Culture; Evaluation of Events, (3) Quality Management (QM), (4) Probabilistic Risk Assessment (PSA), (5) Human Behaviour from the Viewpoint of Industrial Psychology, (6) Technical tour of the Halden Project Experimental Facilities, (7) Human Factors in Control Room Design, (8) Computerised Operator Support Systems (COSSs) and (9) Artificial Intelligence; a new Approach. Most of the contributions are overhead figures from spoken lectures.

Effects of burnup on fission product release and implications for severe fuel damage events

International Nuclear Information System (INIS)

Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

1984-01-01

Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800 0 K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400 0 K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products

Preclinical evaluation of spatial frequency domain-enabled wide-field quantitative imaging for enhanced glioma resection

Science.gov (United States)

Sibai, Mira; Fisher, Carl; Veilleux, Israel; Elliott, Jonathan T.; Leblond, Frederic; Roberts, David W.; Wilson, Brian C.

2017-07-01

5-Aminolevelunic acid-induced protoporphyrin IX (PpIX) fluorescence-guided resection (FGR) enables maximum safe resection of glioma by providing real-time tumor contrast. However, the subjective visual assessment and the variable intrinsic optical attenuation of tissue limit this technique to reliably delineating only high-grade tumors that display strong fluorescence. We have previously shown, using a fiber-optic probe, that quantitative assessment using noninvasive point spectroscopic measurements of the absolute PpIX concentration in tissue further improves the accuracy of FGR, extending it to surgically curable low-grade glioma. More recently, we have shown that implementing spatial frequency domain imaging with a fluorescent-light transport model enables recovery of two-dimensional images of [PpIX], alleviating the need for time-consuming point sampling of the brain surface. We present first results of this technique modified for in vivo imaging on an RG2 rat brain tumor model. Despite the moderate errors in retrieving the absorption and reduced scattering coefficients in the subdiffusive regime of 14% and 19%, respectively, the recovered [PpIX] maps agree within 10% of the point [PpIX] values measured by the fiber-optic probe, validating its potential as an extension or an alternative to point sampling during glioma resection.

Placental Dysfunction Underlies Increased Risk of Fetal Growth Restriction and Stillbirth in Advanced Maternal Age Women.

Science.gov (United States)

Lean, Samantha C; Heazell, Alexander E P; Dilworth, Mark R; Mills, Tracey A; Jones, Rebecca L

2017-08-29

Pregnancies in women of advanced maternal age (AMA) are susceptible to fetal growth restriction (FGR) and stillbirth. We hypothesised that maternal ageing is associated with utero-placental dysfunction, predisposing to adverse fetal outcomes. Women of AMA (≥35 years) and young controls (20-30 years) with uncomplicated pregnancies were studied. Placentas from AMA women exhibited increased syncytial nuclear aggregates and decreased proliferation, and had increased amino acid transporter activity. Chorionic plate and myometrial artery relaxation was increased compared to controls. AMA was associated with lower maternal serum PAPP-A and sFlt and a higher PlGF:sFlt ratio. AMA mice (38-41 weeks) at E17.5 had fewer pups, more late fetal deaths, reduced fetal weight, increased placental weight and reduced fetal:placental weight ratio compared to 8-12 week controls. Maternofetal clearance of 14 C-MeAIB and 3 H-taurine was reduced and uterine arteries showed increased relaxation. These studies identify reduced placental efficiency and altered placental function with AMA in women, with evidence of placental adaptations in normal pregnancies. The AMA mouse model complements the human studies, demonstrating high rates of adverse fetal outcomes and commonalities in placental phenotype. These findings highlight placental dysfunction as a potential mechanism for susceptibility to FGR and stillbirth with AMA.

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

Science.gov (United States)

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

2018-03-13

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

Design and Evaluation of Human System Interfaces (HSIs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

In the safe operation of nuclear power plants and other complex process industries the performance of the control room crews plays an important role. In this respect a well-functioning and well-designed Human-System Interface (HSI) is crucial for safe and efficient operation of the plant. It is therefore essential that the design, development and evaluation of both control rooms and HSI-solutions are conducted in a well-structured way, applying sound human factors principles and guidelines in all phases of the HSI development process. Many nuclear power plants around the world are currently facing major modernisation of their control rooms. In this process computerised, screen-based HSIs replace old conventional operator interfaces. In new control rooms, both in the nuclear field and in other process industries, fully digital, screen-based control rooms are becoming the standard. It is therefore of particular importance to address the design and evaluation of screen-based HSIs in a systematic and consistent way in order to arrive at solutions which take proper advantage of the possibilities for improving operator support through the use of digital, screen-based HSIs, at the same time avoiding pitfalls and problems in the use of this technology. The Halden Reactor Project, in cooperation with the OECD Nuclear Energy Agency, organised an International Summer School on ''Design and Evaluation of Human-System Interfaces (HSIs)'' in Halden, Norway in the period August 25th - 29th, 2003. The Summer School addressed the different steps in design, development and evaluation of HSIs, and the human factors principles, standards and guidelines which should be followed in this process. The lectures comprised both theoretical background, as well as examples of good and bad HSI design, thereby providing practical advice in design and evaluation of operator interfaces and control room solutions to the participants in the Summer School. This CD contains the

Design and Evaluation of Human System Interfaces (HSIs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-07-01

In the safe operation of nuclear power plants and other complex process industries the performance of the control room crews plays an important role. In this respect a well-functioning and well-designed Human-System Interface (HSI) is crucial for safe and efficient operation of the plant. It is therefore essential that the design, development and evaluation of both control rooms and HSI-solutions are conducted in a well-structured way, applying sound human factors principles and guidelines in all phases of the HSI development process. Many nuclear power plants around the world are currently facing major modernisation of their control rooms. In this process computerised, screen-based HSIs replace old conventional operator interfaces. In new control rooms, both in the nuclear field and in other process industries, fully digital, screen-based control rooms are becoming the standard. It is therefore of particular importance to address the design and evaluation of screen-based HSIs in a systematic and consistent way in order to arrive at solutions which take proper advantage of the possibilities for improving operator support through the use of digital, screen-based HSIs, at the same time avoiding pitfalls and problems in the use of this technology. The Halden Reactor Project, in cooperation with the OECD Nuclear Energy Agency, organised an International Summer School on ''Design and Evaluation of Human-System Interfaces (HSIs)'' in Halden, Norway in the period August 25th - 29th, 2003. The Summer School addressed the different steps in design, development and evaluation of HSIs, and the human factors principles, standards and guidelines which should be followed in this process. The lectures comprised both theoretical background, as well as examples of good and bad HSI design, thereby providing practical advice in design and evaluation of operator interfaces and control room solutions to the participants in the Summer School. This CD contains the Proceedings of the

Design and Evaluation of Human System Interfaces (HSIs)

International Nuclear Information System (INIS)

2003-01-01

In the safe operation of nuclear power plants and other complex process industries the performance of the control room crews plays an important role. In this respect a well-functioning and well-designed Human-System Interface (HSI) is crucial for safe and efficient operation of the plant. It is therefore essential that the design, development and evaluation of both control rooms and HSI-solutions are conducted in a well-structured way, applying sound human factors principles and guidelines in all phases of the HSI development process. Many nuclear power plants around the world are currently facing major modernisation of their control rooms. In this process computerised, screen-based HSIs replace old conventional operator interfaces. In new control rooms, both in the nuclear field and in other process industries, fully digital, screen-based control rooms are becoming the standard. It is therefore of particular importance to address the design and evaluation of screen-based HSIs in a systematic and consistent way in order to arrive at solutions which take proper advantage of the possibilities for improving operator support through the use of digital, screen-based HSIs, at the same time avoiding pitfalls and problems in the use of this technology. The Halden Reactor Project, in cooperation with the OECD Nuclear Energy Agency, organised an International Summer School on ''Design and Evaluation of Human-System Interfaces (HSIs)'' in Halden, Norway in the period August 25th - 29th, 2003. The Summer School addressed the different steps in design, development and evaluation of HSIs, and the human factors principles, standards and guidelines which should be followed in this process. The lectures comprised both theoretical background, as well as examples of good and bad HSI design, thereby providing practical advice in design and evaluation of operator interfaces and control room solutions to the participants in the Summer School. This CD contains the Proceedings of the

Development of 10 kV 4H-SiC JBS diode with FGR termination

International Nuclear Information System (INIS)

Huang Runhua; Tao Yonghong; Cao Pengfei; Wang Ling; Li Rui; Chen Gang; Bai Song; Li Yun; Zhao Zhifei

2014-01-01

The design, fabrication, and electrical characteristics of the 4H-SiC JBS diode with a breakdown voltage higher than 10 kV are presented. 60 floating guard rings have been used in the fabrication. Numerical simulations have been performed to select the doping level and thickness of the drift layer and the effectiveness of the edge termination technique. The n-type epilayer is 100 μm in thickness with a doping of 6 × 10 14 cm −3 . The on-state voltage was 2.7 V at J F = 13 A/cm 2 . (semiconductor devices)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Evaluation of strategies for end storage of high-level reactor fuel; Vurdering av strategier for sluttlagring av hoeyaktivt reaktorbrensel

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized.

Evaluation of strategies for end storage of high-level reactor fuel

International Nuclear Information System (INIS)

2001-01-01

This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized

Fission gas release in LWR fuel measured during nuclear operation

International Nuclear Information System (INIS)

Appelhans, A.D.; Skattum, E.; Osetek, D.J.

1980-01-01

A series of fuel behavior experiments are being conducted in the Heavy Boiling Water Reactor in Halden, Norway, to measure the release of Xe, Kr, and I fission products from typical light water reactor design fuel pellets. Helium gas is used to sweep the Xe and Kr fission gases out of two of the Instrumented Fuel Assembly 430 fuel rods and to a gamma spectrometer. The measurements of Xe and Kr are made during nuclear operation at steady state power, and for 135 I following reactor scram. The first experiments were conducted at a burnup of 3000 MWd/t UO 2 , at bulk average fuel temperatures of approx. 850 K and approx. 23 kW/m rod power. The measured release-to-birth ratios (R/B) of Xe and Kr are of the same magnitude as those observed in small UO 2 specimen experiments, when normalized to the estimated fuel surface-to-volume ratio. Preliminary analysis indicates that the release-to-birth ratios can be calculated, using diffusion coefficients determined from small specimen data, to within a factor of approx. 2 for the IFA-430 fuel. The release rate of 135 I is shown to be approximately equal to that of 135 Xe

Chemical Blistering: Cellular and Macromolecular Components

Science.gov (United States)

1985-12-15

34’ has been applied. In principle , this experimental technique involves infection ef a permissive host with a virus which is nonvirulent because of...rate of lactate &cowmulation as compared with the control. The ethanol/methyleno chloride mixture Itself appeared to account for a small amount of...D.C., Part III, pp. 479- 518. 6. Ross, W. C. J. (1962) Biologiral Alkylating Agents* Fundmental ChemLsLr and Desi 2f L ds fgr Selive ToxicitX

Myosin helical pitch angle as a quantitative imaging biomarker for characterization of cardiac programming in fetal growth restriction measured by polarization second harmonic microscopy

Science.gov (United States)

Amat-Roldan, I.; Psilodimitrakopoulos, S.,; Eixarch, E.,; Torre, I.; Wotjas, B.; Crispi, F.; Figueras, F.; Artigas, D.,; Loza-Alvarez, P.; Gratacos, E.,

2009-07-01

Fetal growth restriction (FGR) has recently shown a strong association with cardiac programming which predisposes to cardiovascular mortality in adulthood. Polarization Second Harmonic Microscopy can quantify molecular architecture changes with high sensitivity in cardiac myofibrils. In this work, we use myosin helical pitch angle as an example to quantify such alterations related to this high risk population. Importantly, this shows a potential use of the technique as an early diagnostic tool and an alternative method to understand pathophysiological processes.

Evaluation of neutron irradiation effect on SCC crack growth behaviour for austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

Feedwater flow measurements: challenges, current solutions, and 'soft' developments

International Nuclear Information System (INIS)

Ruan, D.; Roverso, D.; Fantoni, P.F.; Sanabrias, J.I.; Carrasco, J.A.; Fernandez, L.

2002-07-01

This report presents an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of feedwater flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross correlation flow meter concepts. In this report, Section 2 outlines relevant aspects of thermal power calculations on electrical power plants. Section 3 reviews from the available literature possible approaches and solutions for feedwater flow measurement, including ultrasonic flow meters, cross-correlation flow meters, and 'Virtural' flow meters with artificial neural networks. Section 4 reports typical experimental measurements at the Tecnatom's facility. Section 5 presents an integration approach and preliminary experimental tests. Section 6 discusses the role of soft computing techniques in the context of feedwater flow measurements related nuclear fields, and Section 7 highlights the future research direction. (Author)

The biggest bio-fuel plant in Norway - a profitable environmental investment

International Nuclear Information System (INIS)

Lind, Oddvar

2002-01-01

A few years ago, Norske Skog Saugbrugs in Halden, Norway, invested NOK 180 mill in a new combustion plant for bio-fuel. In 2001, the plant produced 400 GWh and so replaced about 35 000 tonnes of oil. Considering the Kyoto Agreement, the profitability is even greater. The capacity of the boiler is 400 - 450 GWh, which covers more than 40 percent of the paper factory's need for thermal energy. The paper factory in Halden is one of the largest in Europe. About half of the bio-fuel derives from the factory's own production, which is an important reason why the price of bio-energy is less than the price of oil. At the same time the use of the biomass for energy production implies that bark and mud does not pile up in the factory. The remaining half of the biomass, the external half, is wood returned from building activities in the form of wood chippings, one-time pallets and similar. This also solves a social problem. The bio-fuel plant uses a fluidized bed boiler of very high efficiency. This implies very small emissions of NOx and CO. Particles are removed by means of an electro filter. The system meets the requirements made by the EU and by Norwegian pollution control authorities

Characterization of reactor coolant by XRF

Energy Technology Data Exchange (ETDEWEB)

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

2002-07-01

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

Characterization of reactor coolant by XRF

International Nuclear Information System (INIS)

Legreid, G.; Beverskog, B.

2002-01-01

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

A survey on the development of advanced instrumentation and control system in NPP

International Nuclear Information System (INIS)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: 1) N4 of EDF in France, 2) AP 600 of Westinghouse in USA, 3) NUPLEX-80+ of ABB-CE in USA, 4) CANDU in Canada, 5) Ohi 3 and 4, APWR and ABWR in Japan, 6) Belt-D in Germany, 7) Sizewell B in Britain, 8) Halden Reactor Projector in Norway, 9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author)

A survey on the development of advanced instrumentation and control system in NPP

Energy Technology Data Exchange (ETDEWEB)

Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1993-12-01

Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: (1) N4 of EDF in France, (2) AP 600 of Westinghouse in USA, (3) NUPLEX-80+ of ABB-CE in USA, (4) CANDU in Canada, (5) Ohi 3 and 4, APWR and ABWR in Japan, (6) Belt-D in Germany, (7) Sizewell B in Britain, (8) Halden Reactor Projector in Norway, (9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author).

On the thermal evolution of Pu-rich agglomerates in MOX

International Nuclear Information System (INIS)

Verwerft, M.; Leenaers, A.; Lippens, M.; Mertens, L.

1999-01-01

From the experience accumulated so far on irradiated MOX fuel, its overall behaviour under irradiation is generally well predicted by existing fuel models. It appears however that additional data are still welcome to properly benchmark fission gas release models, mainly at elevated burnup. To this aim, an international research project, FIGARO, was initiated. Its goal was to provide thermal and fission gas release data og MOX at high burnup. Two MOX fuel rods irradiated to high burnup (50 GWd/tM peak pellet) but at lower power (less than 200 W/cm) were selected for segmentation and instrumentation with central thermocouple and pressure gauge. The instrumented segments were subjected to irradiations at variable linear power in the HALDEN MTR. Both temperature and internal pressure were online monitored during the ramp test. Afterwards, the rod segments were transported and extensively investigated. The paper focuses on the investigation of the evolution of the microstructure of Pu-rich agglomerates as a function of temperature

Modification in the FUDA computer code to predict fuel performance at high burnup

Energy Technology Data Exchange (ETDEWEB)

Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

1997-08-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

Modification in the FUDA computer code to predict fuel performance at high burnup

International Nuclear Information System (INIS)

Das, M.; Arunakumar, B.V.; Prasad, P.N.

1997-01-01

The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

The influence of material variables on corrosion and deuterium uptake of Zr-2.5Nb alloy during irradiation

Energy Technology Data Exchange (ETDEWEB)

McDougall, G.M.; Urbanic, V.F. [Chalk River Technicians Lab., AECL, Ontario (Canada)

2002-07-01

Current CANDU 2 reactors use Zr-2.5Nb pressure tubes that are extruded at 1088 K, cold-drawn 27%, and autoclaved at 673 K for 24 h. This results in a metastable, two-phase microstructure consisting of elongated {alpha}-Zr grains surrounded by a network of {beta}-Zr filaments. To develop a mathematical model of corrosion and deuterium ingress in pressure tubes, we have considered the impact of variables including: fast neutron flux, temperature, and the as fabricated microstructure and its evolution during irradiation. Small specimens of Zr-2.5Nb are being exposed under CANDU water chemistry conditions in the Halden Boiling Water Reactor. The experiments involve fast neutron fluxes (E {>=} 1.05 MeV) of 0, 1.7, and 4.5 x 10{sup 17} n x m{sup -2} - s{sup -1}, and temperatures of 523 and 598 K. Specimens have been prepared from pressure tube materials representative of all current CANDU reactors, materials subject to thermal decomposition of the {beta}-Zr phase, and tubes extruded over a range of conditions. Results from the first three years of the Halden test program are summarized. At both 523 and 598 K, tubes made of {beta}-quenched material exhibit lower oxidation rates than those made from non-{beta}-quenched materials. In short-term out-of-flux exposures at 523 K, three non-{beta}-quenched tubes appear to show linear oxidation kinetics. Similar behavior is not observed in tests conducted out-of-flux at 598 K, or in-flux at either temperature. At 598 K, {beta}-quenched tubes exhibit significantly lower deuterium pickup rates than non-{beta}-quenched tubes. When tested at 598 K, thermally aged specimens show declining oxidation and deuterium pickup rates with increasing {beta}-Zr phase decomposition. At 523 K, the impact of thermal aging was less significant. Preliminary results from an 'extrusion variable test' suggest that tubes fabricated according to the current CANDU specification show the best corrosion resistance. (authors)

Re-fabrication and Instrumentation - resume and outlook

Energy Technology Data Exchange (ETDEWEB)

Kleeman, Hans-Joerg; Oberlaender, Barbara C.

2005-01-01

The special technique ''re-fabrication and instrumentation'' applied at the Institute for Energy Technology (IFE) makes further testing and measuring of irradiated fuel rods in the Halden Boiling Water Reactor (HBWR) possible. Machines, necessary for such operations, were designed and produced at IFE in the years 1991-92. Hot Lab-operations called ''re-fabrication'' include all modifications necessary to load an irradiated fuel rod, commercial or experimental, into the Halden reactor for further testing. ''Instrumentation'' includes all operations necessary to fit instruments into an irradiated fuel rod segment to measure for instance the temperature in the centreline of the fuel stack, pressure increase in the fuel rod and/or changes of the rod length during reactor experiments. The machines used are referred to as: ''Cutting and Grinding Unit'' for machining of the canning; ''Freezing and Drilling Unit'' for drilling of the centreline thermocouple hole in the active fuel stack;''Welding and Drying Unit'' a TIG welding machine for the circumferential welds and the seal welding; and a hydraulic ''Encapsulation Bench''. The ''Welding and Drilling Unit'' includes also a ''He-leak test chamber'' and a ''Hydraulic Press''. In addition, different types of PIE methods are used in quality assurance and documenting of the work done with the fuel rod. For example neutron radiography, visual inspection, dimension measurement, free volume measurements, equipment to measure the flow capability through the fuel stack and different function tests of the measuring devices. This paper will give an overview of the work done so far with this equipment and give information about new equipment which will be installed soon to deal with even more complicated test fuel designs. (Author)

Session summaries for workshop meeting on virtual reality applications in process industry maintenance training, outage planning, control room retrofits and design, 17th - 18th September 1998

International Nuclear Information System (INIS)

Louka, Michael N.

1998-09-01

A well-attended workshop was held in Halden 17th - 18th September 1998 to discuss VR applications in the process industry. In particular, maintenance training, outage planning, decommissioning, control room retrofits, and design were discussed. It is clear that there is a great deal of interest in both current and potential use of VR technology. The workshop participants represented a diverse range of research disciplines, as well as utilities, vendors and regulators (author) (ml)

Proceedings of the specialist meeting on operator aids for severe accidents management and training (SAMOA)

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-07-01

The SAMOA meeting, held in Halden (Norway) in 1993, presented 17 papers grouped into three sessions which titles are: operator aids for control rooms, operator aids for technical support centers, simulation tools for operator training. The specialist meeting also addressed the question of identification of information needs not covered by the instrumentation, examined means to perform phenomenological behaviour assessments needed to support station procedures, and discussed computational aids/methods for predicting accident progression and consequences

Proceedings of the specialist meeting on operator aids for severe accidents management and training (SAMOA)

International Nuclear Information System (INIS)

1993-01-01

The SAMOA meeting, held in Halden (Norway) in 1993, presented 17 papers grouped into three sessions which titles are: operator aids for control rooms, operator aids for technical support centers, simulation tools for operator training. The specialist meeting also addressed the question of identification of information needs not covered by the instrumentation, examined means to perform phenomenological behaviour assessments needed to support station procedures, and discussed computational aids/methods for predicting accident progression and consequences

NEA international co-operative projects

International Nuclear Information System (INIS)

1989-01-01

This text is consecrated at the international co-operative projects of the OECD Nuclear Energy Agency (NEA) in the field of reactor safety (Halden reactor project, Loft project, studies on the damaged Three Mile Island unit-2 reactor, inspection of reactor steel components, incident reporting system) and in the field of radioactive waste management (Stripa project, geochemical data bases, Alligator river project, seabed disposal of high-level radioactive waste, decommissioning of nuclear facilities)

Effects of Sildenafil Citrate and Heparin Treatments on Placental Cell Morphology in a Murine Model of Pregnancy Loss.

Science.gov (United States)

Luna, Rayana Leal; Vasconcelos, Anne Gabrielle; Nunes, Ana Karolina Santana; de Oliveira, Wilma Helena; Barbosa, Karla Patricia de Sousa; Peixoto, Christina Alves

2016-01-01

Lipopolysaccharide (LPS) injections during pregnancy are well established as models for pregnancy complications, including fetal growth restriction (FGR), thrombophilia, preterm labor and abortion. Indeed, inflammation, as induced by LPS injection has been described as a pivotal factor in cases of miscarriage related to placental tissue damage. The phosphodiesterase-5 inhibitor sildenafil (Viagra®) is currently used to treat FGR cases in women, while low-molecular weight heparin (Fragmin®) is a standard treatment for recurrent miscarriage (RM). However, the pathways and cellular dynamics involved in RM are not completely understood. The aim of this study was to evaluate the protective effect of sildenafil and dalteparin in a mouse model of LPS-induced abortion. Histopathology, ultrastructural analysis and immunofluorescence for P-selectin were studied in two different placental cell types: trophoblast cells and labyrinth endothelial cells. Treatment with sildenafil either alone or in combination with heparin showed the best response against LPS-induced injury during pregnancy. In conclusion, our results support the use of these drugs as future therapeutic agents that may protect the placenta against inflammatory injury in RM events. Analyses of the ultrastructure and placental immunophysiology are important to understand the mechanism underlying RM. These findings may spark future studies and aid in the development of new therapies in cases of RM. © 2016 S. Karger AG, Basel.

Xenon migration in UO{sub 2} under irradiation studied by SIMS profilometry

Energy Technology Data Exchange (ETDEWEB)

Marchand, B. [Université de Lyon, CNRS/IN2P3, Université Lyon 1, Institut de Physique Nucléaire de Lyon, 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); AREVA, AREVA NP, 10 rue Juliette Récamier, F-69456 Lyon (France); Moncoffre, N. [Université de Lyon, CNRS/IN2P3, Université Lyon 1, Institut de Physique Nucléaire de Lyon, 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Pipon, Y., E-mail: pipon@ipnl.in2p3.fr [Université de Lyon, CNRS/IN2P3, Université Lyon 1, Institut de Physique Nucléaire de Lyon, 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Université de Lyon, Université Lyon 1, IUT Lyon 1, 43 bd du 11 novembre 1918, 69 622 Villeurbanne cedex (France); Bérerd, N. [Université de Lyon, CNRS/IN2P3, Université Lyon 1, Institut de Physique Nucléaire de Lyon, 4 rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Université de Lyon, Université Lyon 1, IUT Lyon 1, 43 bd du 11 novembre 1918, 69 622 Villeurbanne cedex (France); Garnier, C. [AREVA, AREVA NP, 10 rue Juliette Récamier, F-69456 Lyon (France); Raimbault, L. [Ecole des Mines de Paris, Centre de Géosciences, 35 rue Saint Honoré, F-77305 Fontainebleau cedex (France); Sainsot, P. [Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259, F-69621 Villeurbanne cedex (France); and others

2013-09-15

During Pressurized Water Reactor operation, around 25% of the created Fission Products (FP) are Xenon and Krypton. They have a low solubility in the nuclear fuel and can either (i) agglomerate into bubbles which induce mechanical stress in the fuel pellets or (ii) be released from the pellets, increasing the pressure within the cladding and decreasing the thermal conductivity of the gap between pellets and cladding. After fifty years of studies on the nuclear fuel, all mechanisms of Fission Gas Release (FGR) are still not fully understood. This paper aims at studying the FGR mechanisms by decoupling thermal and irradiation effects and by assessing the Xenon behavior for the first time by profilometry. Samples are first implanted with {sup 136}Xe at 800 keV corresponding to a projected range of 140 nm. They are then either annealed in the temperature range 1400–1600 °C, or irradiated with heavy energy ions (182 MeV Iodine) at Room Temperature (RT), 600 °C or 1000 °C. Depth profiles of implanted Xenon in UO{sub 2} are determined by Secondary Ion Mass Spectrometry (SIMS). It is shown that Xenon is mobile during irradiation at 1000 °C. In contrast, thermal treatments do not induce any Xenon migration process: these results are correlated to the formation of Xenon bubbles observed by Transmission Electron Microscopy.

A fuel performance code TRUST VIc and its validation

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

A fuel performance code TRUST VIc and its validation

International Nuclear Information System (INIS)

Ishida, M.; Kogai, T.

1997-01-01

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

Investigations of time-dependent water pollution effects of overburden dumps - partial project 1. Final report; Untersuchungen gewaesserrelevanter Einfluesse von Bergbauhalden in Abhaengigkeit von der Standzeit - Teilprojekt 1. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Beuge, P.; Dunger, C.; Mibus, J.; Starke, R.

1998-02-01

Management of overburden dumps necessitates careful stocktaking and precise definitions of pollutant emissions so that, if necessary, conclusions can be drawn concerning sanitation concepts in consideration of emission limits and environmental balances. In view of the large number of overburden dumps in Saxony, Saxony-Anhalt and Thuringia caused the authors to restrict their investigations to ore mines and to select elements characteristic of the worked deposits and their potential environmental effects. The following problems were investigated: Are overburden dumps significant emitters of elements at all? Which elements are emitted in what concentrations, in what form and up to what distance from the dump? Are there really different patterns for gangue and impregnation types? How big is the real reaction space of the dump? Should the whole dump be assumed to be a reservoir of elements for emission or just parts of it? What are the factors that most strongly influence element mobilisation and/or retention? What measures should be recommended for emission reduction or prevention from the aspects of necessity, economic efficiency and long-term stability? (orig./SR) [Deutsch] Es ergibt sich fuer den Umgang mit Halden die Notwendigkeit einer gruendlichen Bilanzierung und schaerferen Definition von Schadstoffemission aus ihnen, um daraus Konsequenzen fuer eventuell erforderliche Sanierungskonzeptionen unter Beruecksichtigung von Grenzwerteinhaltung und Umweltbilanz abzuleiten. Der Umfang der einzelnen Fragestellungen und die Bedeutung der Halden in Sachsen, Sachsen-Anhalt und Thueringen sowie Anzahl und ihr Inhalt fuehrte zur Beschraenkung auf den Erzbergbau und auf die Kontrolle ausgewaehlter Elemente, die fuer den jeweils abgebauten Lagerstaetteninhalt charakteristisch und fuer die Wirkung auf die Umwelt von Bedeutung sind. Folgende konkrete Fragestellungen sollen verfolgt werden: - Sind Halden ueberhaupt nennenswerte Emittenten von Elementen? - Welche Elemente werden

Fuel analysis code FAIR and its high burnup modelling capabilities

International Nuclear Information System (INIS)

Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1995-01-01

A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

NKS/SOS-1 seminar on quality assurance

International Nuclear Information System (INIS)

Hammar, L.; Wahlstroem, B.

2001-02-01

The backgrounds and the conduct of the seminar is described. Summaries are given of all presentations and slides shown are appended. An account is given of discussions on different quality issues which were conducted during the seminar in separate groups. Concluding remarks made by the Chairman of NKS are reproduced. Further conclusions will be presented in the main report from the project 'Views on quality assurance at Finnish and Swedish nuclear power plants and the Halden reactor'. (au)

Demonstration of a Robust Sensor System for Remote Condition Monitoring of Heat-Distribution System Manholes

Science.gov (United States)

2016-02-01

The HTP -900RE is capable of transferring Ethernet data over a distance of up to 15 miles with a clear line of sight, and is programmable through a...may be connected to either side of the circuit. The RTU is connected to the FGR- HTP -900-RE radio by a short Ethernet patch cable. The radio is...operations contractor was able to manually query a test pit and read temperatures from through the wireless Ethernet RTU. ERDC/CERL TR-16-2 20 2.3.2

Advanced Joining of Aerospace Metallic Materials.

Science.gov (United States)

1986-07-01

REPAIRED F100 TURBINE VANES tUNDER SIMULATED) SERVICE. CONDI rIONS by A.,I.A.Mom. N.M.Madhava. G.A.KooI and M.IDean 21 REPAIR TECHNIQU ES FOR GAS lII... vapeurs d~gag~es par la decomposition des corps exog~nes. POe) Les cavi t~s (cavit)(fgr4, sont le propre du soudage des fortes paisseurs. Ellas...the most practical information with respect to mechanical properties for gas turbine applications. Generally, sound laser welds (i.e. free from

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

International Nuclear Information System (INIS)

Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

2011-01-01

The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

Nuclear Fuel Design Technology Development for the Future Fuel

Energy Technology Data Exchange (ETDEWEB)

Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

2006-04-15

The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

A questionnaire comparison of two alarm systems

International Nuclear Information System (INIS)

Collier, Steven G.

1997-11-01

A questionnaire was developed, based on guidelines for alarm system design given in NUREG/CR-6105. The intentions were both to develop a subjective instrument for rating the effectiveness of alarm systems and to learn lessons on alarm system design from a comparison of two systems. The questionnaire was administered to reactor operations staff at two locations with different alarm systems embedded in a simulation of the same underlying PWR power plant: Loviisa NPP and Halden Man-Machine Laboratory. The questionnaire, considered as a measuring instrument, had good to high reliability and moderate to good content validity. The questionnaire is considered suitable for further use in the shortened form resulting from this study. Further work is also recommended. The degree of reliability and validity also lend a degree of validation to the NUREG guidelines. The questionnaire was able to show differences between ratings of the two alarm systems. The Loviisa system showed more consistency with other control room features and was better at drawing the operators' attention to important alarms. Both systems were not rated particularly well on alarm prioritisation and spurious alarms. The Halden system was better at showing naturally occurring relationships between alarms. Some of these differences may have been due to the subjects' greater familiarity with the Loviisa alarm system. The results nevertheless show that the questionnaire can measure subjective responses to alarm systems. (author)

Adaptations in Maternofetal Calcium Transport in Relation to Placental Size and Fetal Sex in Mice

Directory of Open Access Journals (Sweden)

Christina E. Hayward

2017-12-01

Full Text Available Appropriate placental transport of calcium is essential for normal fetal skeletal mineralization. In fetal growth restriction (FGR, the failure of a fetus to achieve its growth potential, a number of placental nutrient transport systems show reduced activity but, in the case of calcium, placental transport is increased. In a genetic mouse model of FGR this increase, or adaptation, maintains appropriate fetal calcium content, relative to the size of the fetus, despite a small, dysfunctional placenta. It is unknown whether such an adaptation is also apparent in small, but normally functioning placentas. We tested the hypothesis that calcium transfer would be up-regulated in the lightest vs. heaviest placentas in the same C57Bl/6J wild-type (WT mouse litter. Since lightest placentas are often from females, we also assessed whether fetal sex influenced placental calcium transfer. Placentas and fetuses were collected at embryonic day (E16.5 and 18.5; the lightest and heaviest placentas, and female and male fetuses, were identified. Unidirectional maternofetal calcium clearance (CaKmf was assessed following 45Ca administration to the dam and subsequent radiolabel counts within the fetuses. Placental expression of calcium pathway components was measured by Western blot. Data (median are lightest placenta expressed as percentage of the heaviest within a litter and analyzed by Wilcoxon signed-rank test. In WT mice having normally grown fetuses, CaKmf, per gram placenta near term, in the lightest placentas was increased (126%; P < 0.05 in association with reduced fetal calcium accretion earlier in gestation (92%; P < 0.05, that was subsequently normalized near term. Increased placental expression of calbindin-D9K, an important calcium binding protein, was observed in the lightest placentas near term (122%; P < 0.01. There was no difference in fetal calcium accretion between male and female littermates but a trend toward higher CaKmf in females (P = 0

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

International Nuclear Information System (INIS)

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

2013-01-01

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

NKS/SOS-1 seminar on quality assurance; NKS/SOS-1 seminarium om kvalitetssaekring

Energy Technology Data Exchange (ETDEWEB)

Hammar, L. [ES-Konsult (Sweden); Wahlstroem, B. [VTT Automation (Finland)

2001-02-01

The backgrounds and the conduct of the seminar is described. Summaries are given of all presentations and slides shown are appended. An account is given of discussions on different quality issues which were conducted during the seminar in separate groups. Concluding remarks made by the Chairman of NKS are reproduced. Further conclusions will be presented in the main report from the project 'Views on quality assurance at Finnish and Swedish nuclear power plants and the Halden reactor'. (au)

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, January--March 1976

Energy Technology Data Exchange (ETDEWEB)

Zane, J. O.; Farman, R. F.; Hanson, D. J.; Peterson, A. C.; Ybarrondo, L. J.; Berta, V. T.; Naff, S. A.; Crocker, J. G.; Martinson, Z. R.; Smolik, G. R.; Cawood, G. W.; Quapp, W. J.; Ramsthaler, J. H.; Ransom, V. H.; Scofield, M. P.; Dearien, J. A.; Bohn, M. P.; Burnham, B. W.; James, S. W.; Lee, W. H.; Lime, J. F.; Nalezny, C. L.; MacDonald, P. E.; Thompson, L. B.; Domenico, W. F.; Rice, R. E.; Hendrix, C. E.; Davis, C. B.

1976-06-01

Light water reactor sfaety research performed January through March 1976 is summarized. Results of the Semiscale Mod-1 blowdown heat transfer test series relating to those phenomena that influence core fluid and heat transfer effects are analyzed, and preliminary analyses of the recently completed reflood heat transfer test series are summarized for the forced and gravity feed reflood tests. The first nonnuclear LOCE in the LOFT program was successfully completed and preliminary results are presented. Preliminary results are given for the PCM 8-1 RF Test, the PCM-2A Test, and the Irradiation Effects Scoping Test 2 in the Thermal Fuel Behavior Program. Model development and verification efforts reported in the Reactor Behavior Program include checkout of RELAP4/MOD5 Update 1, development of a new hydrodynamic model for two-phase separated flows, development of the RACHET code to assess the assumptions in current fuel behavior codes of uniform stress and strain in the cladding, modifications of the containment code BEACON, analysis of results from the Halden Assembly IFA-429 helium sorption experiment, development of correlations for the thermal conductivity of UO/sub 2/ and (U,Pu)O/sub 2/, and evaluation of RALAP4 through comparison of calculated results with data from the GE Blowdown Heat Transfer and Semiscale experiments.

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

Energy Technology Data Exchange (ETDEWEB)

Williamson, R. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spencer, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Casagranda, A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, C. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, W. [ANATECH Corp., San Diego, CA (United States); Veearaghavan, S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gardner, R. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to design a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.

Summary of BISON Development and Validation Activities - NEAMS FY16 Report

International Nuclear Information System (INIS)

Williamson, R. L.; Pastore, G.; Gamble, K. A.; Spencer, B. W.; Casagranda, A.; Folsom, C. P.; Liu, W.; Veearaghavan, S.; Novascone, S. R.; Gardner, R. J.; Hales, J. D.

2016-01-01

This summary report contains an overview of work performed under the work package en- titled “FY2016 NEAMS INL-Engineering Scale Fuel Performance (BISON)” A first chapter identifies the specific FY-16 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments, which in- clude: 1) publication of a journal article on solution verification and validation of BISON for LWR fuel, 2) publication of a journal article on 3D Missing Pellet Surface (MPS) analysis of BWR fuel, 3) use of BISON to design a unique 3D MPS validation experiment for future in- stallation in the Halden research reactor, 4) participation in an OECD benchmark on Pellet Clad Mechanical Interaction (PCMI), 5) participation in an OECD benchmark on Reactivity Insertion Accident (RIA) analysis, 6) participation in an OECD activity on uncertainity quantification and sensitivity analysis in nuclear fuel modeling and 7) major improvements to BISON’s fission gas behavior models. A final chapter outlines FY-17 future work.

Formal methods and their applicability in the development of safety critical software systems

International Nuclear Information System (INIS)

Sievertsen, T.

1995-01-01

The OECD Halden Reactor Project has for a number of years been involved in the development and application of a formal software specification and development method based on algebraic specification and the HRP Prover. In parallel to this activity the Project has been evaluating and comparing different methods and approaches to formal software development by their application on realistic case examples. Recent work has demonstrated that algebraic specification and the HRP Prover can be used both in the specification and design of a software system, even down to a concrete model which can be translated into the chosen implementation language. The HRP Prover is currently being used in a case study on the applicability of the methodology in the development of a power range monitoring system for a nuclear power plant. The presentation reviews some of the experiences drawn from the Project's research activities in this area, with special emphasis on questions relating to applicability and limitations, and the role of formal methods in the development of safety-critical software systems. (14 refs., 1 fig.)

Development of oxygen sensors for use in liquid metal

International Nuclear Information System (INIS)

Van Nieuwenhove, Rudi; Ejenstam, Jesper; Szakalos, Peter

2015-01-01

For generation IV reactor concepts, based on liquid metal cooling, there is a need for robust oxygen sensors which can be used in the core of the reactor since corrosion can only be kept sufficiently low by controlling the dissolved oxygen content in the liquid metal. A robust, ceramic membrane type sensor has been developed at IFE/Halden (Norway) and tested in an autoclave system at KTH (Sweden). The sensor has been tested in lead-bismuth at 550 deg. C and performed well. (authors)

Development of oxygen sensors for use in liquid metal

Energy Technology Data Exchange (ETDEWEB)

Van Nieuwenhove, Rudi [Institutt for Energiteknikk, Halden, (Norway); Ejenstam, Jesper; Szakalos, Peter [KTH Royal Institute of Technology, Division of Surface and Corrosion Science, Stockholm, (Sweden)

2015-07-01

For generation IV reactor concepts, based on liquid metal cooling, there is a need for robust oxygen sensors which can be used in the core of the reactor since corrosion can only be kept sufficiently low by controlling the dissolved oxygen content in the liquid metal. A robust, ceramic membrane type sensor has been developed at IFE/Halden (Norway) and tested in an autoclave system at KTH (Sweden). The sensor has been tested in lead-bismuth at 550 deg. C and performed well. (authors)

The OEEC European Nuclear Energy Agency

International Nuclear Information System (INIS)

1961-01-01

The European Nuclear Energy Agency (ENEA) was set up in December 1957 as part of the OEEC to develop nuclear collaboration in Western Europe. The promotion of joint undertakings is one of the most important functions of ENEA, and why one of the first committees of the Agency to be set up was its Top Level Group on Co-operation in the Reactor Field. International collaboration in joint undertakings enables resources in effort, equipment and money to be pooled for the maximum benefit of the countries participating, and is the only way whereby a sufficiently wide range of research possibilities can be covered in a reasonable time. Examples fro such projects are: 1) Halden project - a joint three-year project to exploit the boiling heavy water reactor built by the Norwegian Institute for Atom energy at Halden; 2) Dragon Project - to investigate the possibilities of high-temperature gas-cooled reactors centered on the construction and operation, by an international team, of an experimental 20 MWt high-temperature gas-cooled reactor (Dragon) at the UK Atomic Energy Establishment at Winfrith; 3) Eurochemic - with a principle objective to construct an experimental plant for the treatment of used uranium fuel from reactors in the participating countries; 4) Nuclear Shops. In addition to promoting joint undertakings, a function of ENEA is to encourage scientific and technical collaboration between national research organizations. Co-operation has been facilitated in the areas od nuclear data, food irradiation, environment radioactivity, training, information and nuclear legislation

HAMMLAB 2000 - Long-Term Perspectives for Use of HAMMLAB

International Nuclear Information System (INIS)

Kvalem, J.; Berg, Oe.; Foerdestroemmen, N.T.; Groven, A.-K.; Hollnagel, E.; Pettersen, F.; Solie, Aa.S.; Stokke, E.; Sundling, C.-V.

1996-01-01

This report discusses the perspectives around future use of the Halden Man-Machine Laboratory, HAMMLAB, to ensure that it continues to be the major tool for the Halden Project's research in man-machine systems and process control. The past few years HAMMLAB has undergone major upgrades to fulfill the requirements from the research programmes, and the present HAMMLAB will, with some modifications, be able to serve the research programme for 1997-99. There is, however, a need to look beyond the coming three-year period in order to make sure that HAMMLAB also in the next decades will serve as a global centre of excellence for human-technology interaction studies. Hence, an attempt is made to identify future research needs regarding process control for the beginning of the 21st century. Different aspects of process control are discussed, and their needs for experimental evaluation are pointed out. Based upon the expressed needs for an experimental facility, a discussion is raised around what this means for HAMMLAB, and which specific requirements this imposes on the laboratory. The conclusion is that there is a need for a new, larger and multi-purpose HAMMLAB facility to meet the identified requirements to a complete experimental facility. In the last part of the report a development plan for a new HAMMLAB facility is proposed, and a brief discussion is raised regarding utilisation of resources and the financing aspect. (author)

Modelling isothermal fission gas release

International Nuclear Information System (INIS)

Uffelen, P. van

2002-01-01

The present paper presents a new fission gas release model consisting of two coupled modules. The first module treats the behaviour of the fission gas atoms in spherical grains with a distribution of grain sizes. This module considers single atom diffusion, trapping and fission induced re-solution of gas atoms associated with intragranular bubbles, and re-solution from the grain boundary into a few layers adjacent to the grain face. The second module considers the transport of the fission gas atoms along the grain boundaries. Four mechanisms are incorporated: diffusion controlled precipitation of gas atoms into bubbles, grain boundary bubble sweeping, re-solution of gas atoms into the adjacent grains and gas flow through open porosity when grain boundary bubbles are interconnected. The interconnection of the intergranular bubbles is affected both by the fraction of the grain face occupied by the cavities and by the balance between the bubble internal pressure and the hydrostatic pressure surrounding the bubbles. The model is under validation. In a first step, some numerical routines have been tested by means of analytic solutions. In a second step, the fission gas release model has been coupled with the FTEMP2 code of the Halden Reactor Project for the temperature distribution in the pellets. A parametric study of some steady-state irradiations and one power ramp have been simulated successfully. In particular, the Halden threshold for fission gas release and two simplified FUMEX cases have been computed and are summarised. (author)

The OEEC European Nuclear Energy Agency

Energy Technology Data Exchange (ETDEWEB)

NONE

1961-07-15

The European Nuclear Energy Agency (ENEA) was set up in December 1957 as part of the OEEC to develop nuclear collaboration in Western Europe. The promotion of joint undertakings is one of the most important functions of ENEA, and why one of the first committees of the Agency to be set up was its Top Level Group on Co-operation in the Reactor Field. International collaboration in joint undertakings enables resources in effort, equipment and money to be pooled for the maximum benefit of the countries participating, and is the only way whereby a sufficiently wide range of research possibilities can be covered in a reasonable time. Examples fro such projects are: 1) Halden project - a joint three-year project to exploit the boiling heavy water reactor built by the Norwegian Institute for Atom energy at Halden; 2) Dragon Project - to investigate the possibilities of high-temperature gas-cooled reactors centered on the construction and operation, by an international team, of an experimental 20 MWt high-temperature gas-cooled reactor (Dragon) at the UK Atomic Energy Establishment at Winfrith; 3) Eurochemic - with a principle objective to construct an experimental plant for the treatment of used uranium fuel from reactors in the participating countries; 4) Nuclear Shops. In addition to promoting joint undertakings, a function of ENEA is to encourage scientific and technical collaboration between national research organizations. Co-operation has been facilitated in the areas od nuclear data, food irradiation, environment radioactivity, training, information and nuclear legislation.

Modelling of Rod No 8 in IFA-597:3

International Nuclear Information System (INIS)

Malen, K.

2002-06-01

A Westinghouse Atom 8x8 fuel rod irradiated in the Ringhals 1 BWR for 12 years to a local burnup of about 67 MWd/kgU was refabricated, instrumented with centreline thermocouple and pressure transducer, and irradiated in IFA-597.2 for about 20 days and in IFA-597.3 for about four months. The rod was then sent to Kjeller for puncturing and then to the Studsvik hot cells for detailed post-irradiation examinations. The peak centreline, temperature was close to 1350 deg C. The total fission gas release (FGR) determined from the puncturing was approximately 20 %. Electron probe microanalysis on a fuel section from the central part of the rod showed that virtually 100 % Xe release had occurred in the central part of the pellet out to about half the pellet radius, and this thermal release from the central part of the fuel accounted for the measured total FGR. Optical and scanning electron microscopy of the fuel cross-section showed complete pellet-clad bonding as well as an extensive high burnup 'rim' structure extending at least 0,15 mm in from the fuel surface. The fuel microstructure was characterised at different radial positions in the pellet. This report describes modelling of the rod behaviour using the code SKIROD, in particular fuel temperature and fission gas release. The transient response of the fuel centre line temperature after a scram is also modelled using the code TOODEE2. The modelling results are compared to the experimental results

A novel mouse Fgfr2 mutant, hobbyhorse (hob, exhibits complete XY gonadal sex reversal.

Directory of Open Access Journals (Sweden)

Pam Siggers

Full Text Available The secreted molecule fibroblast growth factor 9 (FGF9 plays a critical role in testis determination in the mouse. In embryonic gonadal somatic cells it is required for maintenance of SOX9 expression, a key determinant of Sertoli cell fate. Conditional gene targeting studies have identified FGFR2 as the main gonadal receptor for FGF9 during sex determination. However, such studies can be complicated by inefficient and variable deletion of floxed alleles, depending on the choice of Cre deleter strain. Here, we report a novel, constitutive allele of Fgfr2, hobbyhorse (hob, which was identified in an ENU-based forward genetic screen for novel testis-determining loci. Fgr2hob is caused by a C to T mutation in the invariant exon 7, resulting in a polypeptide with a mis-sense mutation at position 263 (Pro263Ser in the third extracellular immunoglobulin-like domain of FGFR2. Mutant homozygous embryos show severe limb and lung defects and, when on the sensitised C57BL/6J (B6 genetic background, undergo complete XY gonadal sex reversal associated with failure to maintain expression of Sox9. Genetic crosses employing a null mutant of Fgfr2 suggest that Fgr2hob is a hypomorphic allele, affecting both the FGFR2b and FGFR2c splice isoforms of the receptor. We exploited the consistent phenotype of this constitutive mutant by analysing MAPK signalling at the sex-determining stage of gonad development, but no significant abnormalities in mutant embryos were detected.

Final irradiation and postirradiation data from the NRC/PNL instrumented assembly IFA-432

International Nuclear Information System (INIS)

Lanning, D.D.; Bradley, E.R.

1986-02-01

The instrumented six-rod test assembly IFA-432, sponsored by USNRC, operated in the Halden Reactor from December 1975 to May 1984, with surviving fuel thermocouples, neutron detectors and pressure transducers. Peak burnups of 46 MWd/kgM were achieved. Interim destructive examination data were obtained at peak burnups of 24 and 34 MWd/kgM. This paper presents a synopsis of the irradiation histories and postirradiation examination data for the high burnup rods. The postirradiation condition of the rods correlates well with their individual design parameters and operating conditions

Heating a school by means of waste heat from an ice hall

International Nuclear Information System (INIS)

2001-01-01

As the first building in Norway, Gimle school in Halden can be heated by means of a special combination system that gives up waste heat from a nearby ice hall and earth heat. This system will reduce the expenses of the municipality with the equivalent of USD 30 000 per year, or 618 000 kWh. 308 000 kWh comes from the refrigeration plant of the ice hall and 310 000 kWh from the ground. Although the system is both environmentally friendly end energy conserving, financial state support has been refused

Post irradiation examinations cooperation and worldwide utilization of facilities

International Nuclear Information System (INIS)

Karlsson, Mikael

2009-01-01

Status of post irradiation examinations in Studsvik's facilities, cooperation and worldwide utilization of facilities, was described. Studsvik cooperate with irradiation facilities, as Halden, CEA and JAEA, as well as other hot cell facilities (examples, PSI, ITU and NFD) universities (example, the Royal Institute of Technology in Sweden) in order to be able to provide everything asked for by the nuclear community. Worldwide cooperation for effective use of expensive and highly specialized facilities is important, and the necessity of cooperation will be more and more recognized in the future. (author)

Interaction and control in wearable computing

International Nuclear Information System (INIS)

Strand, Ole Morten; Johansen, Paal; Droeivoldsmo, Asgeir; Reigstad, Magnus; Olsen, Asle; Helgar, Stein

2004-03-01

This report presents the status of Halden Virtual Reality Centre (HVRC) work with technological solutions for wearable computing to support operations where interaction and control of wearable information and communication systems for plant floor personnel are of importance. The report describes a framework and system prototype developed for testing technology, usability and applicability of eye movements and speech for controlling wearable equipment while having both hands free. Potentially interesting areas for further development are discussed with regard to the effect they have on the work situation for plant floor personnel using computerised wearable systems. (Author)

Towards the inclusion of open fabrication porosity in a fission gas release model

Energy Technology Data Exchange (ETDEWEB)

Claisse, Antoine, E-mail: claisse@kth.se [KTH Royal Institute of Technology, Reactor Physics, AlbaNova University Centre, 106 91, Stockholm (Sweden); Van Uffelen, Paul [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125, Karlsruhe (Germany)

2015-11-15

A model is proposed for fission product release in oxide fuels that takes into account the open porosity in a mechanistic manner. Its mathematical framework, assumptions and limitations are presented. It is based on the model for open porosity in the sintering process of crystalline solids. More precisely, a grain is represented by a tetrakaidecahedron and the open porosity is represented by a continuous cylinder along the grain edges. It has been integrated in the TRANSURANUS fuel performance code and applied to the first case of the first FUMEX project as well as to neptunium and americium containing pins irradiated during the SUPERFACT experiment and in the JOYO reactor. The results for LWR and FBR fuels are consistent with the experimental data and the predictions of previous empirical models when the thermal mechanisms are the main drivers of the release, even without using a fitting parameter. They also show a different but somewhat expected behaviour when very high porosity fuels are irradiated at a very low burn-up and at low temperature. - Highlights: • We developed a new athermal FGR model based on the porosity. • We present the model, its framework, assumptions and limitations. • We test it out on several irradiation experiments. • Results are comparable to previous models but without using an empirical parameter.

Experimental study of the effect of task priority and coordination strategy on crew performance

International Nuclear Information System (INIS)

Braarud, Per Oeivind; Ludvigsen, Jan Tore

2002-08-01

This report documents the background and the results from the Teamwork and Task Management experiment 2001 (TTM-2001) performed in the HAlden Man-Machine LABoratory (HAMMLAB). The experiment emphasises concepts that are suggested as an alternative to the application of general workload measures, namely (1) task management; how operators plan, prioritise and accomplish their tasks individually, and (2) teamwork; coordination of work within the team. The concepts were operationalised for the experimental study by crews operating in accordance with 4 work styles combined from 2 experimental factors: Task Priority and Coordination Strategy. The results indicate that Task Priority has no effect on the operator's ability to handle plant malfunction, but that it increases operator ability to prioritise between the importance of the process data, and increases subjective performance. The results demonstrate that Coordination Strategies significantly improve crew performance. However, contrary to the expectations, there is no clear evidence that coordination supports the operator's situation understanding. The stable characteristics of teamwork observed across different tasks may indicate that teamwork is performed in a procedural way, and as a strategy to cope with a complex and uncertain situation. The practical lessons learned from the experiment were that the crews managed to learn the work styles with the given training and were able to perform the work style of the experimental conditions. Thus, it is possible to carry out studies of important task management and teamwork issues in HAMMLAB. (Author)

Measurements of operator performance - an experimental setup

International Nuclear Information System (INIS)

Netland, K.

1980-01-01

The human has to be considered as an important element in a process control system, even if the degree of automation is extremely high. Other elements, e.g. computer, displays, etc., can to a large extent be described and quantified. The human (operator), is difficult to describe in a precise way, and it is just as difficult to predict his thinking and acting in a control room environment. Many factors influence his performance, such as: experience, motivation, level of knowledge, training, control environment, job organization, etc. These factors have to a certain degree to be described before guidelines for design of the man-process interfaces and the control room layout can be developed. For decades, the psychological science has obtained knowledge of the human mind and behaviour. This knowledge should have the potential of a positive input on our effort to describe the factors influencing the operator performance. Even if the human is complex, a better understanding of his thinking and acting, and a more precise description of the factors influencing his performance can be obtained. At OECD Halden Reactor Project an experimental set-up for such studies has been developed and implemented in the computer laboratory. The present set-up includes elements as a computer- and display-based control room, a simulator representing a nuclear power plant, training programme for the subjects, and methods for the experiments. Set-up modules allow reconfiguration of experiments. (orig./HP)

Using CASE-tools based on formal methods in real-life system development of distributed systems

International Nuclear Information System (INIS)

Stoelen, Ketil; Karlsen, Tore Willy; Mohn, Peter; Sandmark, Haaakon

1998-03-01

Within the OECD Halden Reactor Project (HRP) the development and application of formal methods to enhance system quality have been prioritised tasks for the last three years per periods. The three year programme 1997-1999 identifies the need to gain experience from applying formal methods in larger real-life system developments. This motivated the initiation of the HRP research activity Integration of Formal Specification in the Development of HAMMLAB 2000 (INT-FS). The principal objective of INT-FS is to experiment with formal methods in system developments connected to HAMMLAB 2000 and thereby gain a better understanding of their suitability to support practical software engineering. In particular, INT-FS will try to measure the effect of formal methods and gain experience in combining formal methods with traditional development techniques. INT-FS was started up in January 1997. This report describes the status of INT-FS by February 1998. The report identifies objectives and plans; it motivates the choice of formal methods, CASE-tool and software process; it motivates and defines metrics for measuring achievement and the effect of formalization. The report also provides preliminary results from an experimental development of a communication manager; it describes the component to be developed and the background of the participants; it offers some provisional statistics and summarises the experiences with methods and tools. The development of the communication manager is the first attempt ever to exploit state-of-the-art CASE-tools for formal methods in practical software engineering at the HRP. (author)

Wood combustion and NOx formation control

International Nuclear Information System (INIS)

Tewksbury, C.

1991-01-01

The control of wood combustion on stoker fed grates for optimum efficiency and the limiting of NO x (oxides of nitrogen) formation are not necessarily contradictory. This paper presents a matrix of air/fuel ratio control options, then discusses simple on-line monitoring techniques and the importance of operator training and alertness. The significance of uniform fuel feed and air distribution is emphasized. The relationships between combustion control and NO x formation are outlined both in theory and as tested. The experience of the McNeil Generating Station (the largest wood-fired, single boiler, stoker grate, utility electric generating station in the world) is used to demonstrate the theoretical principles. It has been observed that NO x emissions firing 100% whole tree chips with moisture contents as low as 40% by weight can be as low as 0.13 lb/MMBtu (MMBtu = 10 6 Btu) while still achieving a boiler efficiency in the range of 68% to 73% (in the high end of the design range) without the use of post-combustion treatment or flue gas recirculation (FGR). Problems of combustion and emissions control at steaming rates other than normal full-load are also examined. 2 figs., 4 tabs

Fuel performance under normal PWR conditions: A review of relevant experimental results and models

Science.gov (United States)

Charles, M.; Lemaignan, C.

1992-06-01

Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

Annual report 1987

International Nuclear Information System (INIS)

1988-01-01

Work at the Institute for Energy Technoology (IFE) comprises both nuclear and non-nuclear activities. The objectives and framework of the nuclear activities are largely determined by governmental authorities. The international Halden project plays an important role with regard to Norwegian preparedness in the field of nuclear safety, and it also provides a think-tank for work on vital international nuclear safety technology (fuel technology, safety and operation). In addition, IFE provides an essential national function, supplying Norwegian hospitals with isotopes for medical uses, produced using the JEEP II reactor at Kjeller. This reactor is also an important tool for basic physics research

The development of an advanced computerised control room

International Nuclear Information System (INIS)

Haugset, K.

1988-01-01

Control room improvements by use of computer technology is a major activity within the OECD Halden Reactor Project. The goal is to improve operational efficiency and safety by supplying the operator with the information relevant for the specific operational situation, assisting him both in identifying plant state, plan operational strategies and implement such plans. The research activity consists of development of specific operator support systems, validation of such systems under realistic conditions and integration under the scope of an advanced control room concept. The work is carried out in close cooperation with the many member organisations. (author) 2 figs., 8 refs

Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

Energy Technology Data Exchange (ETDEWEB)

Lemmens, K., E-mail: klemmens@sckcen.be [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); González-Robles, E.; Kienzler, B. [Karlsruhe Institute of Technology Institute for Nuclear Waste Disposal (KIT-INE), PO Box 3640, D-76021 Karlsruhe (Germany); Curti, E. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Serrano-Purroy, D. [European Commission, DG Joint Research Centre - JRC, Directorate G - Nuclear Safety & Security, Department G.III, PO Box 2340, D-76125 Karlsruhe (Germany); Sureda, R.; Martínez-Torrents, A. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Roth, O. [Studsvik, Nuclear AB, 611 82 Nyköping (Sweden); Slonszki, E. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary); Mennecart, T. [Waste and Disposal Expert Group, Belgian Nuclear Research Centre (SCK-CEN), Boeretang 200, 2400 Mol (Belgium); Günther-Leopold, I. [Laboratory for Waste Management, Nuclear Energy and Safety Dept., Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Hózer, Z. [Magyar Tudományos Akadémia Energiatudományi Kutatóközpont (MTA EK), PO Box 49, H-1525 Budapest (Hungary)

2017-02-15

The instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45–63 GWd/t{sub HM} and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride – bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H{sub 2} atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways. - Highlights: • Leach tests were performed to study the instant release of fission products from high burn-up UO{sub 2} fuels and one MOX fuel. • In these tests, the fission gas release given by the operator was a pessimistic estimator of the iodine and cesium release. • Iodine and cesium release is proportional to linear power rating beyond 200 W cm{sup −1}. • Closure of the fuel-cladding gap at high burn-up slows down the release. • The release rate decreases following an exponential equation.

Current status of international cooperation on nuclear safety research

International Nuclear Information System (INIS)

Katsuragi, Satoru

1984-01-01

JAERI (Japan Atomic Energy Research Institute), as a representative organization in Japan, has been participating in many international cooperations on nuclear safety research. This report reviews the recent achievement and evolution of the international cooperative safety studies. Twelve projects that are based on the agreements between JAERI and foreign organizations are reviewed. As the fuel irradiation studies, the recent achievement of the OECD Halden Reactor Project and the agreement between Pacific Northwest Laboratories, Battelle Memorial Institute, and JAERI are explained. As for the study of reactivity accident, the cooperation of the NSRR (Nuclear Safety Research Reactor) project in Japan with PBF, PNS and PHEBUS projects in the U.S., West Germany and France, respectively, are now in progress. The fuel performance in abnormal transient and the experiment and analysis of severe fuel damage are the new areas of international interest. The OECD/LOFT project and ROSA-4 projects are also explained in connection with the FP source term problem and the analysis codes such as RELAP-5 and TRAC. As the safety studies associated with the downstream of the nuclear fuel cycle, the BEFAST project of IAEA and the ISIRS project of OECD/NEA are shortly reviewed. (Aoki, K.)

Procedures and Practices - Challenges for Decommissioning Management and Teamwork

Energy Technology Data Exchange (ETDEWEB)

Rindahl, G., E-mail: grete.rindahl@hrp.no [Institute for Energy Technology, Halden (Norway)

2013-08-15

The mental and practical approach to a decommissioning project is often not the same at all levels of an organization. Studies indicate that the early establishment of a decommissioning mindset throughout an organization is an important and frequently overlooked process. It is not enough to establish procedures, if practices and mental approaches are overlooked; and for decommissioning projects that are more often than not dominated by one of a kind problem solving, procedure design is challenging, and new requirements are put on communication. Our research considers stakeholder involvement in these processes in the wider sense of the term; however the main stakeholders in focus are regulators and the work force that will perform or lead the tasks related to decommissioning. Issues here treated include: Decommissioning mindset and the manifestation of mindset issues in decommissioning projects, including challenges and prospective solutions; trust building and trust breaking factors in communication and collaboration relevant to transition and decommissioning; new technologies for collaboration and communication and how these may impair or empower participants - experiences from several domains. This paper is based on work done in collaboration with the OECD NEA Halden Reactor Project. (author)

Predictive factors for intrauterine growth restriction.

Science.gov (United States)

Albu, A R; Anca, A F; Horhoianu, V V; Horhoianu, I A

2014-06-15

Reduced fetal growth is seen in about 10% of the pregnancies but only a minority has a pathological background and is known as intrauterine growth restriction or fetal growth restriction (IUGR / FGR). Increased fetal and neonatal mortality and morbidity as well as adult pathologic conditions are often associated to IUGR. Risk factors for IUGR are easy to assess but have poor predictive value. For the diagnostic purpose, biochemical serum markers, ultrasound and Doppler study of uterine and spiral arteries, placental volume and vascularization, first trimester growth pattern are object of assessment today. Modern evaluations propose combined algorithms using these strategies, all with the goal of a better prediction of risk pregnancies.

Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

International Nuclear Information System (INIS)

Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

2008-01-01

The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

High burnup models in computer code fair

Energy Technology Data Exchange (ETDEWEB)

Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)

1997-08-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.

High burnup models in computer code fair

International Nuclear Information System (INIS)

Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.

1997-01-01

An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs

General knowledge structure for diagnosis

International Nuclear Information System (INIS)

Steinar Brendeford, T.

1996-01-01

At the OECD Halden Reactor Project work has been going on for several years in the field of automatic fault diagnosis for nuclear power plants. Continuing this work, studies are now carried out to combine different diagnostic systems within the same framework. The goal is to establish a general knowledge structure for diagnosis applied to a NPP process. Such a consistent and generic storage of knowledge will lighten the task of combining different diagnosis techniques. An integration like this is expected to increase the robustness and widen the scope of the diagnosis. Further, verification of system reliability and on-line explanations of hypotheses can be helped. Last but not least there is a potential in reuse of both specific and generic knowledge. The general knowledge framework is also a prerequisite for a successful integration of computerized operator support systems within the process supervision and control complex. Consistency, verification and reuse are keywords also in this respect. Systems that should be considered for integration are; automatic control, computerized operator procedures, alarm - and alarm filtering, signal validation, diagnosis and condition based maintenance. This paper presents three prototype diagnosis systems developed at the OECD Halden Reactor Project. A software arrangement for process simulation with these three systems attached in parallel is briefly described. The central part of this setup is a 'blackboard' system to be used for representing shared knowledge. Examples of such knowledge representations are included in the paper. The conclusions so far in this line of work are only tentative. The studies of existing methodologies for diagnosis, however, show a potential for several generalizations to be made in knowledge representation and use. (author). 14 refs, 6 figs

The HAMBO BWR simulator of HAMMLAB

International Nuclear Information System (INIS)

Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

2001-02-01

Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

General knowledge structure for diagnosis

Energy Technology Data Exchange (ETDEWEB)

Steinar Brendeford, T [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

1997-12-31

At the OECD Halden Reactor Project work has been going on for several years in the field of automatic fault diagnosis for nuclear power plants. Continuing this work, studies are now carried out to combine different diagnostic systems within the same framework. The goal is to establish a general knowledge structure for diagnosis applied to a NPP process. Such a consistent and generic storage of knowledge will lighten the task of combining different diagnosis techniques. An integration like this is expected to increase the robustness and widen the scope of the diagnosis. Further, verification of system reliability and on-line explanations of hypotheses can be helped. Last but not least there is a potential in reuse of both specific and generic knowledge. The general knowledge framework is also a prerequisite for a successful integration of computerized operator support systems within the process supervision and control complex. Consistency, verification and reuse are keywords also in this respect. Systems that should be considered for integration are; automatic control, computerized operator procedures, alarm - and alarm filtering, signal validation, diagnosis and condition based maintenance. This paper presents three prototype diagnosis systems developed at the OECD Halden Reactor Project. A software arrangement for process simulation with these three systems attached in parallel is briefly described. The central part of this setup is a `blackboard` system to be used for representing shared knowledge. Examples of such knowledge representations are included in the paper. The conclusions so far in this line of work are only tentative. The studies of existing methodologies for diagnosis, however, show a potential for several generalizations to be made in knowledge representation and use. (author). 14 refs, 6 figs.

Associations of maternal circulating 25-hydroxyvitamin D3 concentration with pregnancy and birth outcomes.

Science.gov (United States)

Rodriguez, A; García-Esteban, R; Basterretxea, M; Lertxundi, A; Rodríguez-Bernal, C; Iñiguez, C; Rodriguez-Dehli, C; Tardón, A; Espada, M; Sunyer, J; Morales, E

2015-11-01

To investigate the association of maternal circulating 25-hydroxyvitamin D3 [25(OH)D3] concentration with pregnancy and birth outcomes. Prospective cohort study. Four geographical areas of Spain, 2003-2008. Of 2382 mother-child pairs participating in the INfancia y Medio Ambiente (INMA) Project. Maternal circulating 25(OH)D3 concentration was measured in pregnancy (mean [SD] 13.5 [2.2] weeks of gestation). We tested associations of maternal 25(OH)D3 concentration with pregnancy and birth outcomes. Gestational diabetes mellitus (GDM), preterm delivery, caesarean section, fetal growth restriction (FGR) and small-for-gestational age (SGA), anthropometric birth outcomes including weight, length and head circumference (HC). Overall, 31.8% and 19.7% of women had vitamin D insufficiency [25(OH)D3 20-29.99 ng/ml] and deficiency [25(OH)D3 < 20 ng/ml], respectively. After adjustment, there was no association between maternal 25(OH)D3 concentration and risk of GDM or preterm delivery. Women with sufficient vitamin D [25(OH)D3 ≥ 30 ng/ml] had a decreased risk of caesarean section by obstructed labour compared with women with vitamin D deficiency [relative risk (RR) = 0.60, 95% CI 0.37, 0.97). Offspring of mothers with higher circulating 25(OH)D3 concentration tended to have smaller HC [coefficient (SE) per doubling concentration of 25(OH)D3, -0.10 (0.05), P = 0.038]. No significant associations were found for other birth outcomes. This study did not find any evidence of an association between vitamin D status in pregnancy and GDM, preterm delivery, FGR, SGA and anthropometric birth outcomes. Results suggest that sufficient circulating vitamin D concentration [25(OH)D3 ≥ 30 ng/ml] in pregnancy may reduce the risk of caesarean section by obstructed labour. © 2014 Royal College of Obstetricians and Gynaecologists.

Rapid radiological characterization method based on the use of dose coefficients

International Nuclear Information System (INIS)

Dulama, C.; Toma, Al.; Dobrin, R.; Valeca, M.

2010-01-01

Intervention actions in case of radiological emergencies and exploratory radiological surveys require rapid methods for the evaluation of the range and extent of contamination. When simple and homogeneous radionuclide composition characterize the radioactive contamination, surrogate measurements can be used to reduce the costs implied by laboratory analyses and to speed-up the process of decision support. A dose-rate measurement-based methodology can be used in conjunction with adequate dose coefficients to assess radionuclide inventories and to calculate dose projections for various intervention scenarios. The paper presents the results obtained for dose coefficients in some particular exposure geometries and the methodology used for deriving dose rate guidelines from activity concentration upper levels specified as contamination limits. All calculations were performed by using the commercial software MicroShield from Grove Software Inc. A test case was selected as to meet the conditions from EPA Federal Guidance Report no. 12 (FGR12) concerning the evaluation of dose coefficients for external exposure from contaminated soil and the obtained results were compared to values given in the referred document. The geometries considered as test cases are: contaminated ground surface; - infinite extended homogeneous surface contamination and soil contaminated to a depth of 15 cm. As shown by the results, the values agree within 50% relative difference for most of the cases. The greatest discrepancies were observed for depth contamination simulation and in the case of radionuclides with complicated gamma emission and this is due to the different approach from MicroShield and FGR12. A case study is presented for validation of the methodology, where both dose rate measurements and laboratory analyses were performed on an extended quasi-homogeneous NORM contamination. The dose rate estimations obtained by applying the dose coefficients to the radionuclide concentrations

A support vector machine integrated system for the classification of operation anomalies in nuclear components and systems

International Nuclear Information System (INIS)

Rocco S, Claudio M.; Zio, Enrico

2007-01-01

A support vector machine (SVM) approach to the classification of transients in nuclear power plants is presented. SVM is a machine-learning algorithm that has been successfully used in pattern recognition for cluster analysis. In the present work, single- and multiclass SVM are combined into a hierarchical structure for distinguishing among transients in nuclear systems on the basis of measured data. An example of application of the approach is presented with respect to the classification of anomalies and malfunctions occurring in the feedwater system of a boiling water reactor. The data used in the example are provided by the HAMBO simulator of the Halden Reactor Project

Summary of BISON Development Activities: NEAMS FY14 Report

Energy Technology Data Exchange (ETDEWEB)

Williamson, R. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spencer, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, W. [Anatech, Inc.; Pastore, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Perez, D. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gardner, R. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stafford, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-10-01

This summary report contains an overview of work performed under the work package entitled “FY2014 NEAMS INL-Engineering Scale Fuel Performance & Interface with RPL Tools.” A first chapter identifies the specific FY-14 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes substantial additional work including 1) efforts to improve numerical convergence and contact in BISON, 2) development of capability to simulate hydrogen behavior in Zircaloy cladding and 3) efforts to enhance collaborative work with the Halden Research Program. A final chapter briefly outlines planned future work.

Manifestations of nonlinearity in fuel center thermocouple steady-state and transient data: implications for data analysis

International Nuclear Information System (INIS)

Lanning, D.D.; Barnes, B.O.; Williford, R.E.

1979-01-01

The interpretation and verification of fuel centerline thermocouple data are analyzed. Two new concepts are discussed along with their application to in-reactor data from IFA-432, a heavily instrumented six-rod Halden reactor test assembly sponsored by the Nuclear Regulatory Commission. The main ideas presented in this report are that: it is more useful to plot resistance versus power than simply to plot temperature versus power; and the response of the centerline temperature to a linear power decrease is correlated to the rod's current resistance-vs-power behavior. Thus, the resistance-vs-power measurement can be verified by performing a linear power decrease and by plotting the temperature response

Evolution of international nuclear cooperation

International Nuclear Information System (INIS)

Goldschmidt, Bertrand

1978-01-01

The various stages of the history of the International Cooperation in nuclear matters are reviewed: isolationism period at the end of the war with the anglosaxon monopoly; opening period with the 1955 Geneva Conference and the creation of organisms within the framework of Euratom (the six European countries), of OCED (NEA) and of UNO (IAEA); industrial realizations period with Euratom research centres (Ispra, Geel, Karlsruhe and Petten) and the NEA enterprises (Halden, Dragon project, Eurochemic). The international industrial accords in the domains of exploitation and uranium enrichment are recalled and the program for the international evaluating of the fuel cycle (INFCE) is mentioned [fr

The experimental evaluation of the success path monitoring system

International Nuclear Information System (INIS)

Baker, S.C.; Marshall, E.C.; Reiersen, C.S.; Owre, F.; Gaudio, P.J. Jr.

1988-01-01

The Success Path Monitoring System (SPMS) is an advanced computer-based operator aid which is intended to enhance the operator's ability to handle plant disturbances effectively. It achieves this by providing an on-line assessment of both the status of critical safety functions and the status of appropriate success paths. A prototype version of SPMS has been implemented on the OECD Halden Project's PWR simulator, the objective being to test the SPMS in a realistic situation and to assess whether it performed in accordance with design expectations. 16 reactor operators were observed coping with complex transient scenarios both with, and without SPMS being available

Fission gas release from UO2 pellet fuel at high burn-up

International Nuclear Information System (INIS)

Vitanza, C.; Kolstad, E.; Graziani, U.

1979-01-01

Analysis of in-reactor measurements of fuel center temperature and rod internal pressure at the OECD Halden Reactor Project has led to the development of an empirical fission gas release model, which is described. The model originally derived from data obtained in the low and intermediate burn-up range, appears to give good predictions for rods irradiated to high exposures as well. PIE puncturing data from seven fuel rods, operated at relatively constant powers and peak center temperatures between 1900 and 2000 0 C up to approx. 40,000 MWd/t UO 2 , did not exhibit any burn-up enhancement on the fission gas release rate

Thermal performance of annular-coated and sphere-pac LWR fuel rod designs

International Nuclear Information System (INIS)

Guenther, R.J.; Hsieh, K.A.; Barner, J.O.; Freshley, M.D.

1980-01-01

Two FCI-resistant UO 2 fuel rod designs are being compared to a reference design in irradiation tests in the Halden Boiling Water Reactor (HBWR) as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The primary fuel design (annular-coated-pressurized) incorporates annular pellets, a graphite coating on the inner surface of the Zircaloy cladding, and pressurized helium fill gas. Also being investigated is an 87% smear density sphere-pac design with pressurized helium fill gas. The solid pellet (reference) and annular-coated designs described had helium fill gas at approx. 100 kPa and the sphere-pac rods were pressurized at approx. 455 kPa

Sensitivity assessment of fuel performance codes for LOCA accident scenario

Energy Technology Data Exchange (ETDEWEB)

Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

2017-07-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

YKAe - Research programme on nuclear power plant systems behaviour and operational aspects of safety

International Nuclear Information System (INIS)

Mattila, L.; Vanttola, T.

1992-01-01

The major part of nuclear energy research in Finland has been organised as five-year nationally coordinated research programs. The research programme on Systems Behaviour and Operational Aspects of Safety is under way during 1990-1994. Its annual volume has been about 35 person-years and its annual expenditure about FIM 18 million. Studies in the field on safe operational margins of nuclear fuel and reactor core concentrate on fuel high burn-up behaviour, VVER fuel experiments, and reactor core behaviour in complex reactivity transients such as 3-D phenomena and ATWS events. The PACTEL facility is used for the thermal hydraulic studies of the Loviisa type reactors (scaled 1:305). Validation of accident analysis codes is carried out by participation in international standard problems. Advanced foreign computer codes for severe reactor accidents are implemented, modified as needed and applied to level-2 PSAs and the improvement of accident management procedures. Fire simulation methods are tested using data from experiments in the German HDR facility. A nuclear plant analyzer for efficient safety analyses is being developed using the APROS process simulation environment. Computerized operator support systems are being studied in cooperation with the OECD Halden Project. The basic factors affecting plant operator activities and the development of their competence are being investigated. A comprehensive system for the control of plant operational safety is being developed by combining living PSA and safety indicators

Sensitivity assessment of fuel performance codes for LOCA accident scenario

International Nuclear Information System (INIS)

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

2017-01-01

FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

Forecasters Handbook for Central America and Adjacent Waters

Science.gov (United States)

1989-09-01

8217 VoIlo1 de Chiriqui aeasMa~ ODLE Valcain de Agua 0 J.r7ai-a -,f Cs4, Valcari Santa Ania Z4,, jpw SIERRA BE LAS MINAS eNcy Volcari San Miguel, ia Volcan...Nicaragua NICARAGUA % 0.Los Chiles P Colorado beria e. ,he H~gQ& il~a\\~CSPuerto Viejo 183’W 182’W Nicoy APuerto ere0Pndn la Iaz6RA L Lia CARIBBEAN SEA 10’N...34"" . ,’. ’ . , . - Fgr 4.’--.*..*4 . ici".al Shelf off A -- i L’ , i" T i Nic’a• agua • s... .n/ C o.astal St/tm a t .- 3

IFPA meeting 2015 workshop report II: mechanistic role of the placenta in fetal programming; biomarkers of placental function and complications of pregnancy.

Science.gov (United States)

Andraweera, P H; Bobek, G; Bowen, C; Burton, G J; Correa Frigerio, P; Chaparro, A; Dickinson, H; Duncombe, G; Hyett, J; Illanes, S E; Johnstone, E; Kumar, S; Morgan, T K; Myers, J; Orefice, R; Roberts, C T; Salafia, C M; Thornburg, K L; Whitehead, C L; Bainbridge, S A

2016-12-01

Workshops are an integral component of the annual International Federation of Placenta Association (IFPA) meeting, allowing for networking and focused discussion related to specialized topics on the placenta. At the 2015 IFPA meeting (Brisbane, Australia) twelve themed workshops were held, three of which are summarized in this report. These workshops focused on various aspects of placental function, particularly in cases of placenta-mediated disease. Collectively, these inter-connected workshops highlighted the role of the placenta in fetal programming, the use of various biomarkers to monitor placental function across pregnancy, and the clinical impact of novel diagnostic and surveillance modalities in instances of late onset fetal growth restriction (FGR). Copyright © 2015 Elsevier Ltd. All rights reserved.

Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

Energy Technology Data Exchange (ETDEWEB)

Eberle, R; Heins, L; Sontheimer, F [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-08-01

The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ``on the safe side``. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ``go/no-go`` conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU`s CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs.

Fuel rod analysis to respond to high burnup and demanding loading requirements. Probabilistic methodology recovers design margins narrowed by degrading fuel thermal conductivity and progressing FGR

International Nuclear Information System (INIS)

Eberle, R.; Heins, L.; Sontheimer, F.

1997-01-01

The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ''on the safe side''. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ''go/no-go'' conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU's CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs

The effects of zinc on cobalt deposition in PWRs: summary report

International Nuclear Information System (INIS)

Bennet, Peter

1996-01-01

An experiment has been performed in a PWR loop of the Halden reactor to investigate the effects of the addition of 50 ppb zinc to the coolant on the incorporation of cobalt into the oxide films on primary circuit constructional materials. This report summarises the results from the three phases of the test. It was shown that zinc addition inhibits the corrosion of both new metal surfaces and surfaces with well-established oxides; this results in thinner oxide layers and reduced incorporation of cobalt into the oxide. Generally, there were no significant differences between the deposition of cobalt-60 onto pre-oxidised and new metal surfaces. In Phase 1 of the experiment, Co-60 deposition rates (normalised to the circulating Co-60 concentration) were lower than those measured in previous experiments in the loop by factors in the range from 5 to 10. In Phase 2, differences were observed in the behaviour of iron- and nickel-based alloys: larger decreases in the deposition rate compared with Phase 1 took place for stainless steel samples (i.e. factors > 20), whilst decreases on nickel-based coupons were generally less than a factor of 5. Co-60 deposition rates onto stainless steel coupons newly installed for Phase 3 of the experiment were greater by an order of magnitude than on coupons which had been exposed for all three phases; i.e. they were similar to those observed in Phase 1. The mechanisms by which zinc acts to inhibit corrosion and the incorporation of activity into oxide layers are not fully understood. More experimental data are required to resolve this issue, including information on the chemical form of the zinc within the oxide layer. (author)

Data acquisition. GRAAL experiment. Hybrid reactor experiment. AMS experiment

International Nuclear Information System (INIS)

Barancourt, D.; Barbier, G.; Bosson, G.; Bouvier, J.; Gallin-Martel, L.; Meillon, B.; Stassi, P.; Tournier, M.

1997-01-01

The main activity of the data acquisition team has consisted in hardware and software developments for the GRAAL experiment with the trigger board, for the 'Reacteurs Hybrides' group with an acquisition board ADCVME8V and for the AMS experiment with the monitoring of the aerogel detector. (authors)

Software diversity: way to enhance safety?

International Nuclear Information System (INIS)

Dahll, G.; Bishop, P.

1990-01-01

The topic of the paper is the use of diversely produced programs to enhance the safety of computer-based systems applied in safety-critical areas. The paper starts with a survey of scientific investigations on the impact of software redundancy made at various institutions around the world. Main emphasis will, however, be put on the PODS/STEM projects, which have been performed at the OECD Halden Project in cooperation with the Technical Research Center of Finland, the Safety and Reliability Directorate, AEA Technology, UK, and Central Electricity Research Laboratory (now National Power Technology and Environment Centre), UK. In these projects, three program versions were made independently by three different teams, all based on the same specification. The three programs were tested back-to-back with a large amount of test data. The experience and results from this process were carefully logged and used for further analysis. Various strategies for test data selection were compared, with respect to fault finding strategies, as well as to branch and statement coverages of the tested programs. The assumption of independence of failures in diversely produced programs was investigated. A particularly interesting effect, namely failure masking due to program structure, was revealed. Static analysis techniques, software measures, and software reliability estimates were also studied. (author)

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.

Annual report 1989

International Nuclear Information System (INIS)

1989-01-01

Work at Institutt for Energiteknikk (IFE) comprises both nuclear and non-nuclear activities. Nuclear power research at the Institute is performed within the international OECD Halden Reactor Project. In the current agreement period (1988-1990) more than forty organizations are involved in the project, representing eight European nations, Japan and USA. In 1989 about 45 different fuel element designs have been tested. In the fuel program, work is focussing on the basic factors governing fuel reliability and safety, under normal running conditions and during operational disturbances and accident situations. This involves characterization of the mechanical, thermal and chemical properties of the fuels. In the recent years efforts have increasingly been directed at corrosion, and materials technology in general. The results have been used by the project members as a basis for licensing new fuel element constructions, and also for the verification of analytical/empirical models for fuel reliability. The second main area covered by the Halden Project is information technology, i.e. computer-based monitoring and control systems. The activities in this field centre on an advanced experimental control room coupled to a full-scale simulator of a nuclear power station of the pressurized water type. The laboratory was upgraded in 1989 with the addition of several new full graphics work stations and general software for the development of real-time expert systems. In addition to developing and testing a varity of operator support systems, an important task in 1989 has been to carry out a detailed test program for a computer-based operating procedure system. Full interpretation of the data has not yet been completed, but preliminary results suggest that the automated system provides superior information support to the operator in a wide range of situations

Design and Validation of Control Room Upgrades Using a Research Simulator Facility

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; Vivek Agarwal; Jeffrey C. Joe; Julius J. Persensky

2012-11-01

Since 1981, the United States (U.S.) Nuclear Regulatory Commission (NRC) [1] requires a plant- specific simulator facility for use in training at U.S. nuclear power plants (NPPs). These training simulators are in near constant use for training and qualification of licensed NPP operators. In the early 1980s, the Halden Man-Machine Laboratory (HAMMLab) at the Halden Reactor Project (HRP) in Norway first built perhaps the most well known set of research simulators. The HRP offered a high- fidelity simulator facility in which the simulator is functionally linked to a specific plant but in which the human-machine interface (HMI) may differ from that found in the plant. As such, HAMMLab incorporated more advanced digital instrumentation and controls (I&C) than the plant, thereby giving it considerable interface flexibility that researchers took full advantage of when designing and validating different ways to upgrade NPP control rooms. Several U.S. partners—the U.S. NRC, the Electrical Power Research Institute (EPRI), Sandia National Laboratories, and Idaho National Laboratory (INL) – as well as international members of the HRP, have been working with HRP to run control room simulator studies. These studies, which use crews from Scandinavian plants, are used to determine crew behavior in a variety of normal and off-normal plant operations. The findings have ultimately been used to guide safety considerations at plants and to inform advanced HMI design—both for the regulator and in industry. Given the desire to use U.S. crews of licensed operators on a simulator of a U.S. NPP, there is a clear need for a research simulator facility in the U.S. There is no general-purpose reconfigurable research oriented control room simulator facility in the U.S. that can be used for a variety of studies, including the design and validation of control room upgrades.

The evaluation of failure stress and released amount of fission product gas of power ramped rod by fuel behaviour analysis code 'FEMAXI-III'

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujita, Misao

1984-01-01

Pellet-Cladding Interaction(PCI) related in-pile failure of Zircaloy sheathed fuel rod is in general considered to be caused by combination of pellet-cladding mechanical interaction(PCMI) with fuel-cladding chemical interaction(FCCI). An understanding of a basic mechanism of PCI-related fuel failure is therefore necessary to get actual cladding hoop stress from mechanical interaction and released amounts of fission product(FP) gas of aggressive environmental agency from chemical interaction. This paper describes results of code analysis performed on fuel failure to cladding hoop stress and amounts of FP gas released under the condition associated with power ramping. Data from Halden(HBWR) and from Studsvik(R2) are used for code analysis. The fuel behaviour analysis code ''FEMAXI-III'' is used as an analytical tool. The followings are revealed from the study: (1) PCI-related fuel failure is dependent upon cladding hoop stress and released amounts of FP gas at power ramping. (2) Preliminary calculated threshold values of hoop stress and of released amounts of FP gas to PCI failure are respectively 330MPa, 10% under the Halden condition, 190MPa, 5% under the Inter ramp(BWR) condition, and 270MPa, 14% under the Over ramp(PWR) condition. The values of hoop stress calculated are almost in the similar range of those obtained from ex-reactor PCI simulated tests searched from references published. (3) The FEMAXI-III code verification is made in mechanical manner by using in-pile deformation data(diametral strain) obtained from power ramping test undertaken by JAERI. While, the code verification is made in thermal manner by using punctured FP gas data obtained from post irradiation examination performed on non-defected power ramped fuel rods. The calculations are resulted in good agreements to both, mechanical and thermal experimental data suggesting the validity of the code evaluation. (J.P.N.)

Annual report 1985 - IFE

International Nuclear Information System (INIS)

At present Norwegian nuclear energy research is centered around the international OECD Halden Reactor Project which has participants from 40 organisations in 10 countries, including the Nordic nations and most major nuclear power countries within OECD. The research programme is concentrated on nuclear fuel and safety technology and on computer-based methods for operation and supervision of power reactors. Fuel research is now being concentrated more and more on characterizing long-term effects with regard to the efficiency and reliablility of the fuel. Special instruments developed in the Halden project are essential for implementing the fuel testing programme. These instruments provide detailed information concerning external and internal conditions in the fuel rods. In turn such data provide a basis for developing calculation models for use in design, for licensing and when determining operating strategies for the different types of fuel. Work to develop a large computer-based system for assisting operators under irregular operating conditions was commenced in 1985. This system will consist of several modules that help trace and diagnose disturbances, and it contains procedures for returning the system to an acceptable state. A knowledge-based system making use of modern techniques for storing alarm sequences and a series of important parameters for the various operating faults, is being developed to help in diagnosing and recognizing fault situations. The nuclear supervision system SCORPIO has been supplied for all three pressurized water reactors in the Swedish nuclear power plant at Ringhals. Operation of the three sets of plant can now be planned from one central workstation. Cooperation with Ringhals is continuing with a view to connecting SCORPIO direct to the plant, so that the operating margins can be monitored in true time at a detailed level. Other signatory nations are displaying an interest in this monitoring system, especially West Germany and USA

Fuel rod behaviour during transients

International Nuclear Information System (INIS)

Bilsby, C.F.; Haste, T.J.; Garlick, A.; Cameron, R.F.

1982-04-01

The clad deformation code CANSWELL-2 is described. This is used, either as a stand-alone code or within MABEL-2, to predict and analyse the results of LOCA simulations in the Halden and NRU reactors and in the KfK and PROPAT rigs. Experimental evidence on fuel behaviour in RIA, PCM and ATWS events is presented with inclusion of certain FRAP-T5 results. Published calculations from the accident codes FRAP-T4 and FRAP-T5 are compared with experimental results in simulated loss of coolant tests in the Power Burst Facility. The limitations of this code in its treatment of RIA, PCM and ATWS events are considered. (U.K.)

NEA activities in 1993. 22. Annual Report of the OECD Nuclear Energy Agency

International Nuclear Information System (INIS)

1994-01-01

The titles and themes of the ten chapters of this report on NEA activities are: trends in nuclear energy; nuclear development and the fuel cycle (potential contribution of nuclear energy, policy alternatives, maintaining the nuclear option, prospective); reactor safety and regulation (safety research, regulatory approach, safety assessment, accident phenomenology and management, human factors, international standards); radiation protection (revision of the standards, assessment of the protection, international emergency exercises); radioactive waste management (long term safety assessment, in situ evaluation, other radioactive wastes); nuclear science (role, nuclear data, use of supercomputers, actinide transmutation, NEA Data Bank); joint projects (Three Mile Island vessel investigation, Halden reactor project...); legal affairs (liability aspects...); information programme; relations with non-member countries. 28 figs

Institute for Atomic Energy

International Nuclear Information System (INIS)

1980-01-01

The Institute has in 1980 changed its name to 'Institutt for Energiteknikk' and this reflects a de facto change in programme emphasis in which a large part of the activity in 1979 was in energy systems analysis, energy technology and conservation, petroleum technology,etc. However previous projects in environmental and safety aspects of nuclear power, risk analysis and fundamental physics using neutron beams, have continued. Nuclear technology is now concentrated in the Halden Reactor Project, whose work is outlined. Isotope production based on the JEEP II reactor and irradiation there and in the Co-60 plant, and isotope applications in environmental and resource investigations continue as previously. Waste processing and safeguards are also carried out as national responsibilities. (JIW)

Environmentally assisted cracking in LWR materials

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Park, J.H.; Shack, W.J.; Zhang, J.; Brust, F.W.; Dong, P.

1998-01-01

The effect of dissolved oxygen level on fatigue life of austenitic stainless steels is discussed and the results of a detailed study of the effect of the environment on the growth of cracks during fatigue initiation are presented. Initial test results are given for specimens irradiated in the Halden reactor. Impurities introduced by shielded metal arc welding that may affect susceptibility to stress corrosion cracking are described. Results of calculations of residual stresses in core shroud weldments are summarized. Crack growth rates of high-nickel alloys under cyclic loading with R ratios from 0.2--0.95 in water that contains a wide range of dissolved oxygen and hydrogen concentrations at 289 and 320 C are summarized

Overview of the PBF test results

International Nuclear Information System (INIS)

Zeile, H.J.

1980-01-01

The Thermal Fuels Behavior Program (TFBP) of EG and G Idaho conducts fuel behavior research in the Power Burst Facility (PBF) at INEL and at the Halden Reactor in Norway. The fuels behavior research in the PBF is directed toward providing a detailed understanding of the response of light water reactor (LWR) nuclear fuel assemblies to off-normal and hypothesized accident conditions. Single fuel rods and clusters of highly instrumented fuel rods are installed within a central test space of the PBF core for testing. The core can be operated in various modes to provide test conditions typical of accidents and off-normal conditions that may be experienced in a pressurized water reactor or a boiling water reactor

Development and evaluation of a function-oriented display system: background and initial results

International Nuclear Information System (INIS)

Andresen, Gisle; Pirus, Dominique

2005-01-01

Screen-based Human System Interfaces (HSI) are gradually replacing the conventional panel-based HSls, although no clear design philosophy for screen-based HSIs exists. The current paper presents a comprehensive design philosophy where a function-analysis of the plant forms the backbone of the information requirements, information presentation and display organization. The main characteristics of the concept are described as well as the development process behind the first prototype. Findings from the first usability test of the prototype are reported and potential benefits of the HSI are discussed. The work is part of OECD Halden Reactor Project's ongoing research on innovative design for advanced NPP control-rooms and is conducted in close co-operation with Electricite de France

Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes

International Nuclear Information System (INIS)

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

2013-01-01

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

Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

The Experiment Factory: standardizing behavioral experiments

Directory of Open Access Journals (Sweden)

Vanessa V Sochat

2016-04-01

Full Text Available The administration of behavioral and experimental paradigms for psychology research is hindered by lack of a coordinated effort to develop and deploy standardized paradigms. While several frameworks (de Leeuw (2015; McDonnell et al. (2012; Mason and Suri (2011; Lange et al. (2015 have provided infrastructure and methods for individual research groups to develop paradigms, missing is a coordinated effort to develop paradigms linked with a system to easily deploy them. This disorganization leads to redundancy in development, divergent implementations of conceptually identical tasks, disorganized and error-prone code lacking documentation, and difficulty in replication. The ongoing reproducibility crisis in psychology and neuroscience research (Baker (2015; Open Science Collaboration (2015 highlights the urgency of this challenge: reproducible research in behavioral psychology is conditional on deployment of equivalent experiments. A large, accessible repository of experiments for researchers to develop collaboratively is most efficiently accomplished through an open source framework. Here we present the Experiment Factory, an open source framework for the development and deployment of web-based experiments. The modular infrastructure includes experiments, virtual machines for local or cloud deployment, and an application to drive these components and provide developers with functions and tools for further extension. We release this infrastructure with a deployment (http://www.expfactory.org that researchers are currently using to run a set of over 80 standardized web-based experiments on Amazon Mechanical Turk. By providing open source tools for both deployment and development, this novel infrastructure holds promise to bring reproducibility to the administration of experiments, and accelerate scientific progress by providing a shared community resource of psychological paradigms.

Training in virtual reality: qualitative results from a comparison of technology types

International Nuclear Information System (INIS)

Sebok, Angelia; Nystad, Espen

2005-08-01

The study described in HWR-734 consisted of three experiments. The first experiment compared procedural and configuration learning using four display technology types; the second experiment compared these same four technology types in a retention and transfer of training condition, and the third experiment compared assembly learning using three types of displays. The purpose of the study was to determine if the technology types differed in their ability to support learning. Thus, several different types of learning tasks were included (i.e., procedures, configuration and assembly). For all types of learning, short term training effectiveness was evaluated: subjects were tested immediately after the training session, in the same conditions in which they had been trained. In addition, procedural learning was also evaluated in a retention and transfer of training condition, where subjects were tested 24 hours after the initial training, in a real-world talk-through condition. In the Procedural / Configuration knowledge conditions, four technology types were evaluated. These include a desktop monoscopic display (DM), a desktop stereoscopic display (DS), a large-screen stereoscopic display (LS-S), and a head-mounted display with orientation tracking. In the Assembly knowledge condition, three technology types were evaluated: a desktop monoscopic display (DM), a large-screen monoscopic display (LS-M), and a head-mounted display without orientation tracking (HMD-nt). Twenty-four employees at the Halden Boiling Water Reactor (HBWR) participated in the study. The study lasted for one week, so all subjects attended for multiple days. Prior to starting the experiment, subjects attended a briefing where they were given instructions on how to use the equipment. They were allowed 10 minutes practice to get familiar with each technology type before each experimental session. During and after the experimental sessions, data were collected. These included objective data, such as

Customer experience

OpenAIRE

Koperdáková, Zuzana

2016-01-01

Bachelor thesis deals with the theme of customer experience and terms related to this topic. The thesis consists of three parts. The first part explains the terms generally, as the experience or customer loyalty. The second part is dedicated to medotology used for Customer Experience Management. In the third part is described application of Customer Experience Management in practice, particularly in the context Touch Point Analyses in GE Money Bank.

Fetal umbilical artery Doppler pulsatility index and childhood neurocognitive outcome at 12 years.

Science.gov (United States)

Mone, Fionnuala; McConnell, Barbara; Thompson, Andrew; Segurado, Ricardo; Hepper, Peter; Stewart, Moira C; Dornan, James C; Ong, Stephen; McAuliffe, Fionnuala M; Shields, Michael D

2016-06-15

To determine whether an elevated fetal umbilical artery Doppler (UAD) pulsatility index (PI) at 28 weeks' gestation, in the absence of fetal growth restriction (FGR) and prematurity, is associated with adverse neurocognitive outcome in children aged 12 years. Prospective cohort study, comparing children with a normal fetal UAD PI (<90th centile) (n=110) and those with an elevated PI (≥90th centile) (n=40). UAD was performed at 28, 32 and 34 weeks gestation. At 12 years of age, all children were assessed under standardised conditions at Queen's University, Belfast, UK to determine cognitive and behavioural outcomes using the British Ability Score-II and Achenbach Child Behavioural Checklist Parent Rated Version under standardised conditions. Regression analysis was performed, controlling for confounders such as gender, socioeconomic status and age at assessment. The mean age of follow-up was 12.4 years (±0.5 SD) with 44% of children male (n=63). When UAD was assessed at 28 weeks, the elevated fetal UAD group had lower scores in cognitive assessments of information processing and memory. Parameters included (1) recall of objects immediate verbal (p=0.002), (2) delayed verbal (p=0.008) and (3) recall of objects immediate spatial (p=0.0016). There were no significant differences between the Doppler groups at 32 or 34 weeks' gestation. An elevated UAD PI at 28 weeks' gestation in the absence of FGR or prematurity is associated with lower scores of declarative memory in children aged 12 years. A potential explanation for this is an element of placental insufficiency in the presence of the appropriately grown fetus, which affects the development of the fetal hippocampus and information processing and memory long-term. These findings, however, had no impact on overall academic ability, mental processing and reasoning or overall behavioural function. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a

Presolar silicates in the matrix and fine-grained rims around chondrules in primitive CO3.0 chondrites: Evidence for pre-accretionary aqueous alteration of the rims in the solar nebula

Science.gov (United States)

Haenecour, Pierre; Floss, Christine; Zega, Thomas J.; Croat, Thomas K.; Wang, Alian; Jolliff, Bradley L.; Carpenter, Paul

2018-01-01

To investigate the origin of fine-grained rims around chondrules (FGRs), we compared presolar grain abundances, elemental compositions and mineralogies in fine-grained interstitial matrix material and individual FGRs in the primitive CO3.0 chondrites Allan Hills A77307, LaPaz Icefield 031117 and Dominion Range 08006. The observation of similar overall O-anomalous (∼155 ppm) and C-anomalous grain abundances (∼40 ppm) in all three CO3.0 chondrites suggests that they all accreted from a nebular reservoir with similar presolar grain abundances. The presence of presolar silicate grains in FGRs combined with the observation of similar estimated porosity between interstitial matrix regions and FGRs in LAP 031117 and ALHA77307, as well as the identification of a composite FGR (a small rimmed chondrule within a larger chondrule rim) in ALHA77307, all provide evidence for a formation of FGRs by accretion of dust grains onto freely-floating chondrules in the solar nebula before their aggregation into their parent body asteroids. Our study also shows systematically lower abundances of presolar silicate grains in the FGRs than in the matrix regions of CO3 chondrites, while the abundances of SiC grains are the same in all areas, within errors. This trend differs from CR2 chondrites in which the presolar silicate abundances are higher in the FGRs than in the matrix, but similar to each other within 2σ errors. This observation combined with the identification of localized (micrometer-scaled) aqueous alteration in a FGR of LAP 031117 suggests that the lower abundance of presolar silicates in FGRs reflects pre-accretionary aqueous alteration of the fine-grained material in the FGRs. This pre-accretionary alteration could be due to either hydration and heating of freely floating rimmed chondrules in icy regions of the solar nebula or melted water ice associated with 26Al-related heating inside precursor planetesimals, followed by aggregation of FGRs into the CO chondrite parent-body.

Simulator experiments: effects of NPP operator experience on performance

International Nuclear Information System (INIS)

Beare, A.N.; Gray, L.H.

1984-01-01

During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE Boiling Water Reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of senior reactor operator (SRO) experience, operating crew augmentation with an STA and practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. Sixteen two-man crews of licensed operators were employed in a 2 x 2 factorial design. The SROs leading the crews were split into high and low experience groups on the basis of their years of experience as an SRO. One half of the high- and low-SRO experience groups were assisted by an STA. The crews responded to four simulated plant casualties. A five-variable set of content-referenced performance measures was derived from task analyses of the procedurally correct responses to the four casualties. System parameters and control manipulations were recorded by the computer controlling the simulator. Data on communications and procedure use were obtained from analysis of videotapes of the exercises. Questionnaires were used to collect subject biographical information and data on subjective workload during each simulated casualty. For four of the five performance measures, no significant differences were found between groups led by high (25 to 114 months) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have significantly shorter task performance times than crews led by high experience SROs. The presence of the STA had no significant effect on overall team performance in responding to the four simulated casualties. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator

Dependence of the time-constant of a fuel rod on different design and operational parameters

International Nuclear Information System (INIS)

Elenkov, D.; Lassmann, K.; Schubert, A.; Laar, J. van de

2001-01-01

The temperature response during a reactor shutdown has been measured for many years in the OECD-Halden Project. It has been shown that the complicated shutdown processes can be characterized by a time constant τ which depends on different fuel design and operational parameters, such as fuel geometry, gap size, fill gas pressure and composition, burnup and linear heat rate. In the paper the concept of a time constant is analyzed and the dependence of the time constant on various parameters is investigated analytically. Measured time constants for different designs and conditions are compared with those derived from calculations of the TRANSURANUS code. Employing standard models results in a systematic underprediction of the time constant, i.e. the heat transfer during shutdown is overestimated. (author)

Human-centred radiological software techniques supporting improved nuclear safety

International Nuclear Information System (INIS)

Szoeke, Istvan; Johnsen, Terje

2013-01-01

The Institute for Energy Technology (IFE) is an international research foundation for energy and nuclear technology. IFE is also the host for the international OECD Halden Reactor Project. The Software Engineering Department in the Man Technology Organisation at IFE is a leading international centre of competence for the development and evaluation of human-centred technologies, process visualisation, and the lifecycle of high integrity software important to safety. This paper is an attempt to give a general overview of the current, and some of the foreseen, research and development of human-centred radiological software technologies at the Software Engineering department to meet with the need of improved radiological safety for not only nuclear industry but also other industries around the world. (author)

Proceedings of the topical meeting on reactivity initiated accidents (RIA)

International Nuclear Information System (INIS)

2003-01-01

The topical meeting was devoted to RIA fuel acceptance criteria, in particular to the fuel fragmentation enthalpy limit and the PCMI failure enthalpy limit in relation to high burnup fuel. In total 50 participants attended. Research and industry organisations from France, Finland, Germany, Hungary, Japan, the Russian Federation, Sweden, the United Kingdom and the USA including the Swiss regulatory body, HSK, and the Halden reactor project, presented 16 papers in all. The papers covered three main areas: 'best estimate' core calculations for RIA energy deposition in high burnup fuels, the technical background of current and new RIA fuel safety criteria, and ongoing RIA experimental programmes. A number of open issues were identified, whose resolution is expected from ongoing and planned national and international experimental programmes

4. Activity report of the Nuclear Energy Agency. 1975

International Nuclear Information System (INIS)

1976-01-01

Despite the many economic and related difficulties experienced throughout the OECD area during 1975, for nuclear power the year brought much promise, together with consolidation and some incouragement for the nuclear industry. 1975 saw a concentration of NEA's technical work on nuclear safety, radioactive waste management, and studies related to the nuclear fuel cycle. NEA's work on regulatory questions was also important. Besides NEA's substantial involvement in the preparation of Extension Agreements for the Halden and Dragon Reactor Projects and for the International Food Irradiation Project, as well as the Agreement to establish a research and development program at Eurochemic on high-activity waste treatment, the Agency has also been closely concerned with scientific and technological developments both within these Projects and in other areas

Specific features of the WWER Uranium-Gadolinium fuel behavior at BOL

International Nuclear Information System (INIS)

Shcheglov, A.; Proselkov, V.; Volkov, B.

2013-01-01

The calculated-experimental analysis of the WWER fuel behavior with 5%wt of gadolinium oxide at the beginning of life (BOL) is presented. The results are based on the data on fuel centerline temperature measurements, gas media pressure inside the cladding and fuel elongation obtained during irradiation of the test fuel rods in HBWR (Halden). Computer analysis of experimental data is performed with TOPRA-2, version 2 code. It is shown that specific features of the uranium-gadolinium fuel behavior at the early of life is due to presence of burnable absorber influencing the average linear heat rating, radial power distribution and lower thermal conductivity. In particular, the analysis of “late” relocation effect on the maximum Gd fuel temperature is presented. (authors)

Research and experience report 2014. Developments in the technical and legal areas of nuclear monitoring

International Nuclear Information System (INIS)

2015-04-01

The research into regulatory safety carried out by the Swiss Federal Nuclear Safety Inspectorate (ENSI) serves to develop the tools that ENSI requires for the fulfilment of its responsibilities. The programme is divided into seven areas: 1) research on fuels and materials covers the reactor core and the multiple successive barriers used for the containment of radioactive materials. It is concentrated on high burn-ups and safety criteria. Research into structural materials is focused on ageing mechanisms. The SAFE Project has been investigating the formation and growth of cracks in materials used in reactor cooling circuits. The researchers at the Paul Scherrer Institute (PSI) obtained results on how the hydrogen present in hot water as well as the sequence of mechanical stresses affect crack development; 2) the OECD Projects on internal events and damage encourage international exchange of experience on incidents. Specific data bases facilitate the analysis of relevant operating experience from many countries. The ICDE Project, which is looking at common-cause failures in the components used in nuclear power plants, published an overview report on heat exchangers; 3) ENSI supports research projects on external events such as aircraft crashes, flooding and earthquakes. The Swiss Seismological Service (SED) published a report on ground motion attenuation as a function of increasing distance to the earthquake's source. By incorporating data from countries with high seismic activity, the SED has improved the attenuation model for Switzerland. The international SMART project has been looking at the impact of severe earthquakes on nuclear power plant buildings in order to represent the dynamic behaviour and vulnerability of reinforced concrete structures; 4) as far as human factors are concerned, the Halden Reactor Project completed informative simulation studies. A comparative study of 10 operator groups revealed marked variability in the way unforeseen situations

Extracting Insights from Experience Designers to Enhance User Experience Design

OpenAIRE

Kremer, Simon; Lindemann, Udo

2016-01-01

User Experience (UX) summarizes how a user expects, perceives and assesses an encounter with a product. User Experience Design (UXD) aims at creating meaningful experiences. While UXD is a rather young discipline with-in product development and traditional processes predominate, other disciplines traditionally focus on creating experiences. We engaged with experience de-signers from the fields of arts, movies, sports, music and event management. By analyzing their working processes via interv...

Nondestructive fission gas release measurement and analysis

International Nuclear Information System (INIS)

O'Leary, P.M.; Packard, D.R.

1993-01-01

Siemens Power Corporation (SPC) has performed reactor poolside gamma scanning measurements of fuel rods for fission gas release (FGR) detection for more than 10 yr. The measurement system has been previously described. Over the years, the data acquisition system, the method of spectrum analysis, and the means of reducing spectrum interference have been significantly improved. A personal computer (PC)-based multichannel analyzer (MCA) package is used to collect, display, and store high-resolution gamma-ray spectra measured in the fuel rod plenum. A PC spread sheet is used to fit the measured spectra and compute sample count rates after Compton background subtraction. A Zircaloy plenum spacer is often used to reduce positron annihilation interference that can arise from the INCONEL reg-sign plenum spring used in SPC-manufactured fuel rods

Wire system aging assessment and condition monitoring (WASCO)

International Nuclear Information System (INIS)

Fantoni, P.F.; Nordlund, A.

2006-04-01

Nuclear facilities rely on electrical wire systems to perform a variety of functions for successful operation. Many of these functions directly support the safe operation of the facility; therefore, the continued reliability of wire systems, even as they age, is critical. Condition Monitoring (CM) of installed wire systems is an important part of any aging program, both during the first 40 years of the qualified life and even more in anticipation of the license renewal for a nuclear power plant. This report describes a method for wire system condition monitoring, developed at the Halden Reactor Project, which is based on Frequency Domain Reflectometry. This method resulted in the development of a system called LIRA (LIne Resonance Analysis), which can be used on-line to detect any local or global changes in the cable electrical parameters as a consequence of insulation faults or degradation. LIRA is composed of a signal generator, a signal analyser and a simulator that can be used to simulate several failure/degradation scenarios and assess the accuracy and sensitivity of the LIRA system. Chapter 5 of this report describes an complementary approach based on positron measurement techniques, used widely in defect physics due to the high sensitivity to micro defects, in particular open volume defects. This report describes in details these methodologies, the results of field experiments and the proposed future work. (au)

NEA activities in 1994

International Nuclear Information System (INIS)

1995-01-01

This report deals with the activities in 1994 of the OECD Nuclear Energy Agency including : trends in nuclear power, nuclear development and fuel cycle (data, tools, methodologies, nuclear industry and its context, fuel cycle policy, energy policy reviews), reactor safety and regulation (exchange of operating experience and human factors, accident prevention activities, severe-accident phenomenology and management, structural integrity, probabilistic safety assessment, fuel cycle safety, regulatory approaches to severe-accident issues, inspection practices), radiation protection (radiation protection today and tomorrow, application of radiation protection standards, protection of workers, international emergency exercises), radioactive waste management (the philosophical and ethical basis of disposal, safety assessment and site investigations, other radiation waste management activities), nuclear science (scientific studies, validation of working methods, the NEA data bank, projects in support of other NEA programmes), joint projects (the Rasplav project, the Halden reactor project, information system on occupational exposure, decommissioning of nuclear facilities), legal affairs (liability and nuclear safety assistance to Eastern Europe, training seminar on nuclear law, information on nuclear law), information programme (review of Nea information programme, publications programme, a major information forum), and relations with non m ember countries (NEA relations with non-members countries, the Central and Eastern European Countries and New Independent States of the former Soviet Union programme of co-operation and assistance). (O.L.). 22 figs

Wire system aging assessment and condition monitoring (WASCO)

Energy Technology Data Exchange (ETDEWEB)

Fantoni, P.F. [Institutt for energiteknikk (Norway); Nordlund, A. [Chalmers Univ. of Technology (Sweden)

2006-04-15

Nuclear facilities rely on electrical wire systems to perform a variety of functions for successful operation. Many of these functions directly support the safe operation of the facility; therefore, the continued reliability of wire systems, even as they age, is critical. Condition Monitoring (CM) of installed wire systems is an important part of any aging program, both during the first 40 years of the qualified life and even more in anticipation of the license renewal for a nuclear power plant. This report describes a method for wire system condition monitoring, developed at the Halden Reactor Project, which is based on Frequency Domain Reflectometry. This method resulted in the development of a system called LIRA (LIne Resonance Analysis), which can be used on-line to detect any local or global changes in the cable electrical parameters as a consequence of insulation faults or degradation. LIRA is composed of a signal generator, a signal analyser and a simulator that can be used to simulate several failure/degradation scenarios and assess the accuracy and sensitivity of the LIRA system. Chapter 5 of this report describes an complementary approach based on positron measurement techniques, used widely in defect physics due to the high sensitivity to micro defects, in particular open volume defects. This report describes in details these methodologies, the results of field experiments and the proposed future work. (au)

Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-08-15

Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

The experience sampling method: Investigating students' affective experience

Science.gov (United States)

Nissen, Jayson M.; Stetzer, MacKenzie R.; Shemwell, Jonathan T.

2013-01-01

Improving non-cognitive outcomes such as attitudes, efficacy, and persistence in physics courses is an important goal of physics education. This investigation implemented an in-the-moment surveying technique called the Experience Sampling Method (ESM) [1] to measure students' affective experience in physics. Measurements included: self-efficacy, cognitive efficiency, activation, intrinsic motivation, and affect. Data are presented that show contrasts in students' experiences (e.g., in physics vs. non-physics courses).

Simulator experiments: effects of NPP operator experience on performance

International Nuclear Information System (INIS)

Beare, A.N.; Gray, L.H.

1985-01-01

Experiments are being conducted on nuclear power plant (NPP) control room training simulators by the Oak Ridge National Laboratory, its subcontractor, General Physics Corporation, and participating utilities. The experiments are sponsored by the Nuclear Regulatory Commission's (NRC) Human Factors and Safeguards Branch, Division of Risk Analysis and Operations, and are a continuation of prior research using simulators, supported by field data collection, to provide a technical basis for NRC human factors regulatory issues concerned with the operational safety of nuclear power plants. During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE boiling water reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of (a) senior reactor operator (SRO) experience, (b) operating crew augmentation with an STA and (c) practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator. Methodology and results to date are reported

Experiment WA1 (CDHS Neutrino Experiment)

CERN Multimedia

CERN PhotoLab

1977-01-01

Experiment WA1, also known under CDHS (CERN, Dortmund, Heidelberg, Saclay; spokesman Jack Steinberger), was the first neutrino experiment on the SPS, in its West Area. Magnetized iron (with a toroidal field) forms the core of the detector. On its outside we see drift chambers and photomultipliers (detecting the light from the plastic scintillators further in). Peter Schilly is wearing a white coat. See also CERN Annual Report 1976, p.57.

TRIO experiment

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Malecha, R.F.

1984-09-01

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an anaytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion

Pixel Experiments

DEFF Research Database (Denmark)

Petersen, Kjell Yngve; Søndergaard, Karin; Augustesen, Christina

2015-01-01

Pixel Experiments The term pixel is traditionally defined as any of the minute elements that together constitute a larger context or image. A pixel has its own form and is the smallest unit seen within a larger structure. In working with the potentials of LED technology in architectural lighting...... for using LED lighting in lighting design practice. The speculative experiments that have been set-up have aimed to clarify the variables that can be used as parameters in the design of lighting applications; including, for example, the structuring and software control of light. The experiments also...... elucidate and exemplify already well-known problems in relation to the experience of vertical and horizontal lighting. Pixel Experiments exist as a synergy between speculative test setups and lighting design in practice. This book is one of four books that is published in connection with the research...

ARC EMCS Experiments (Seedling Growth-2) Experiment Status

Science.gov (United States)

Heathcote, David; Steele, Marianne

2015-01-01

Presentation of the status of the ARC ISS (International Space Station) Experiment, Seedling Growth-2 to the Payload Operations Investigator Working Group meeting at MSFC, Huntsville AL. The experiment employs the European Modular Cultivation System (ECMS).

TRACY transient experiment databook. 2) ramp withdrawal experiment

International Nuclear Information System (INIS)

Nakajima, Ken; Yamane, Yuichi; Ogawa, Kazuhiko; Aizawa, Eiju; Yanagisawa, Hiroshi; Miyoshi, Yoshinori

2002-03-01

This is a databook of TRACY ''ramp withdrawal'' experiments. TRACY is a reactor to perform supercritical experiments using low-enriched uranyl nitrate aqueous solution. The excess reactivity of TRACY is 3$ at maximum, and it is inserted by feeding the solution to a core tank or by withdrawing a control rod, which is called as the transient rod, from the core. In the ramp withdrawal experiment, the supercritical experiment is initiated by withdrawing the transient rod from the core in a constant speed using a motor drive system. The data in the present databook consist of datasheets and graphs. Experimental conditions and typical values of measured parameters are tabulated in the datasheet. In the graph, power and temperature profiles are plotted. Those data are useful for the investigation of criticality accidents with fissile solutions, and for validation of criticality accident analysis codes. (author)