Water management of the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Rabusic, P.

1990-01-01

Industrial water for the Dukovany nuclear power plant is taken from a reservoir built on the Jihlava river. The volume of the reservoir is 17.1 million m 3 . The water taken serves mainly as the make-up water for the cooling circuit and is treated by clearing; it is also used for the preparation of demineralized water for making up the condensate circuit and for the preparation of soft water for making up the heat piping. The consumption of industrial water is 1 to 2 m 3 /s. Waste waters are released into rainwater drainage, sewage drainage and industrial water drainage. Waste waters are segregated according to the place of their origin and are purified, mainly on ion exchangers and on an evaporator. In normal conditions, they are returned to the operation. Concentrated liquid residues and solid wastes will be stored by using the multiple containment system. The most important radioisotopes that may be present in nuclear power plant waste waters, the water and radionuclide balance of the Dukovany nuclear power plant, and chemical and radiological data on the Dukovany waste waters (1989) are tabulated. (P.A.). 6 refs

Refurbishment of the NPP Dukovany I and C System

International Nuclear Information System (INIS)

Karpeta, C.; Rosol, J.

2004-01-01

An overview of the NPP Dukovany Instrumentation and Control (I and C) refurbishment project is presented in this paper from the standpoint of both its management and technical aspects. Reasons for taking the decision to replace the original plant I and C system are outlined and the objectives set for the refurbishment project are stated. The paper is focusing on describing more in detail the first part of the refurbishment, i.e. replacement of the I and C portions of the plant systems important to nuclear safety and the process information system. This includes the reactor trip system, engineered safety features actuation system, reactor power limitation system, reactor power control system, post-accident monitoring system, reactor core monitoring system and unit data acquisition and processing system. Information is given on the main processes of the project, i.e. the bidding, design, manufacturing, installation and commissioning. Specific licensing process applied to this refurbishment project is also outlined. An account of the current status of the project implementation is given. (author)

20 Years of Dukovany NPP Operation and Plans for Future

International Nuclear Information System (INIS)

Vlcek, J.

2006-01-01

NPP Dukovany (4x440MW), the first Czech nuclear power plant and a keystone of Czech Electricity Grid, in 2005 celebrated 20 years of successful operation. In my presentation I am going to speak about next areas. I will start about construction and commissioning history of Dukovany NPP, then I'll put forward what milestones of Dukovany NPP operation were accomplished, I'll say examples about our modernisation and investment activities (including replacement of IandC equipment, modifications for increasing of power and support grid services) and in the end our plans for future. (author)

Dukovany nuclear power plant safety

International Nuclear Information System (INIS)

1999-01-01

Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

Dukovany NPP maintenance management

International Nuclear Information System (INIS)

Siegel, F.

2005-01-01

Maintenance planning and management for the technological equipment of the Dukovany NPP are described. A Maintenance Control and Evaluation System has been developed and is in use to help manage the complex maintenance issue. Practical examples of outputs of the System, with a comprehensive use of the cost, reliability and safety related data, are presented. (author)

Computer programme for operator support in primary frequency control of NPP Dukovany

International Nuclear Information System (INIS)

Kurka, J.; Petruzela, I.; Piroutek, Z.

1996-01-01

The computer programme for the control room operator support in the primary frequency control operation of the NPP Dukovany power unit is described. This operational mode has recently been implemented. Until now, the plant has been exclusively operated in the based-load mode at nominal power. The stability of the main unit parameters, unit power output and main steam header pressure, is the characteristic feature for this mode of operation of units with WWER 440 type of reactors. This does no more hold in the new mode in which these both principal parameters vary in dependency on the power network frequency changes. This fact, however, has an important impact on the activities of the control room operators. Therefore, to help them to adapt themselves easily and without unnecessary stresses to new conditions the regulatory body when licensing the new mode finally approved its implementation only if computerized operator support system would be installed in the control room. Such a computer program for the operator support has been recently developed in EGU Prague and has already been implemented into the unit information system on all four Dukovany units. (author). 3 refs, 7 figs

Some results of periodical analysis of the behavior of unit 4 at the Dukovany NPP - II

International Nuclear Information System (INIS)

Jirsa, P.

1993-01-01

The results of a detailed spectral analysis of noise signals recorded at the unit 4 reactor of the Dukovany NPP during and after the start-up stage are presented. The changes in the noise source signal spectra during the start-up and in normal operation were evaluated by multidimensional autoregression analysis. The temperature dependence of the signal spectra peaks from individual detectors is described. The sources of the peaks at 27.9 and 37 Hz during normal operation were identified as standing pressure waves in the primary circuit of the reactor. 3 tabs., 23 figs., 6 refs

IAEA-led Operational Safety Team Reviews Dukovany Nuclear Power Plant, Czech Republic

International Nuclear Information System (INIS)

2011-01-01

indicator for the Collective Effective Dose of Dukovany NPP has shown exceptionally good results in comparison with other Pressurized Water Reactors (PWRs) and Water-Water Energetic Reactors (VVERs). During the past three years, the value was less than 0.15 manSv per Unit; and A comprehensive station blackout simulator scenario has been developed and implemented for several years at the plant. The scenario involves multiple off-site agencies and has a high degree of realism. It is used to practise and improve response times for power restoration activities and familiarize off-site personnel with nuclear safety principles. The team has also made recommendations and suggestions related to areas where the operational safety of the Dukovany NPP could be improved. Examples include: Improving the effectiveness of using its root cause analysis method to prevent the repetition of events; Clarifying and reinforcing expectations regarding the use of error prevention tools and enhancing their application; and Ensuring adequate protection of emergency workers and evacuees following a possible radioactive release. The plant's management expressed a determination to address all the areas identified for improvement and requested that the IAEA schedule a follow-up mission in approximately 18 months. The team handed over a draft of their recommendations, suggestions and good practices to the plant management in the form of ''Technical Notes'' for factual comments. The technical notes will be reviewed at the IAEA headquarters, including any comments from Dukovany NPP and the Czech regulatory body, the State Office for Nuclear Safety (SUJB). The final report will be submitted to the Government of the Czech Republic within three months. This was the 162nd mission of the OSART programme, which began in 1982. Video and photos are available from the OSART Mission. General information about OSART missions can be found on the IAEA Website. (IAEA)

A way of Dukovany NPP to privatization and liberalized market

International Nuclear Information System (INIS)

Kouklik, I.

2001-01-01

This presentation describes the current situation in the Dukovany NPP with two main upcoming future phenomena - privatization of CEZ company and preparation for liberalized electricity market. Considerations about level of safety, investment costs and competitiveness of Dukovany Nuclear Power Plant in liberalization process is described together with a comparison of some safety features of the world's NPPs. Results of this comparison are used for consideration and evaluation of some require modifications effectiveness. (author)

Dukovany NPP operation

International Nuclear Information System (INIS)

Vlcek, Jaroslav

2010-01-01

The topics discussed include: Dukovany NPP among CEZ Group power plants; International missions at the plant; Plant operation results; and Strategic goals and challenges. Historical data are presented in the graphical form, such as the unit capacity factor, unplanned capability loss factor, unplanned automatic scrams, fuel reliability, industrial safety accident rate, collective radiation exposure, WANO index, power generation data, and maximum achievable power by the end of year. Also discussed were the company culture and human resources, maintenance, power uprate, and related phenomena. (P.A.)

From the chronicle of training of Dukovany NPP staff

International Nuclear Information System (INIS)

2005-01-01

The long way the Dukovany NPP had to go before the plant staff was fully qualified and skilled is described. First the training concept was prepared, then the necessary training facilities were set up, lecturers and instructors were hired and trained, training programmes and training materials were developed, and ultimately the first training course was launched in 1979. A training NPP was constructed and a full-scope simulator of the Dukovany NPP was set up. The current status of organization of NPP staff training by the CEZ utility is highlighted. (author)

Upgrade the startup test instrumentation (ANMS) with innovated I and C system at the Dukovany WWER 440/213 units

International Nuclear Information System (INIS)

Kocek, V.

2005-01-01

The paper focuses on the innovation procedures of physics startup tests instrumentation used in Dukovany NPP (ANMS) connected with I and C system refurbishment of the Dukovany WWER 440/213 units. The following matters are further discussed in this paper: a) changes in ANMS design, whose has been carried out due to reflect the situation connected with I and C innovation; b) process of consistency checking during implementation stage; c) results of the physics startup tests with upgraded ANMS in 2005 (introduction of Gd2 fuel on Dukovany NPP) (Authors)

Calculation of fluences of fast neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor. Part I. Theory, calculations, comparison with the experiment

International Nuclear Information System (INIS)

Rataj, J.

1993-10-01

The method of calculating neutron spectra and integral flux densities of neutrons hitting the pressure vessel of the Dukovany NPP WWER-440 reactor is outlined. The one-dimensional and two-dimensional calculations were performed by means of the DORT code in R, R-Z, and R-Θ geometries using the cross sections from the ELXSIR library. In the R-Θ geometry, the coupled neutron flux densities were determined. The calculated values of the maximum activation of detectors differ less than 15% from the values measured in surveillance specimens, which is within the limit of uncertainty associated with the position of the detector in the casing. The differences between the calculated and observed data behind the pressure vessel were below 4%. 10 tabs., 3 figs., 41 refs

Development of the relations between the Dukovany NPPs and the local community at present time

International Nuclear Information System (INIS)

Mozor, B.

2000-01-01

Dukovany NPP is the only one operating in Czech Republic, with 4 PWR units, producing about a quarter of the total electricity supplied by the biggest electric power company in the country. On the same site there is a spent fuel storage with a capacity of 600 tones, and a new spent fuel storage planned. This will have sufficient capacity for storing spent fuel for 40 years of Dukovany NPP operation. The communication of the NPP with the public is based on the following rules: open information, open dialog with target groups, mutual exchange of impartial information on operation, serious and tolerant attitude of the Dukovany NPP bilateral dialog with the surroundings. This resulted in the fact that in Czech republic 50% of of the population is in favour of NPPs, and this percent is even higher in the Dukovany region (80%)

NPP Dukovany modernisation programme for long-term safe and efficient operation

International Nuclear Information System (INIS)

Krizek, K.; Sabata, M.; Vasa, I.

2004-01-01

The strategy, main items and issues, responsibilities and linkages of the modernisation programme are outlined. The long-term operation (LTO) concept has been approved by the power plant board of directors, including the scope of activities and schedule, related processes have been defined, a working group of 40 specialists has been set up to manage the LTO process. The technical and economic optimum for the Dukovany LTO will be determined by the end of 2006, and a Technical LTO Programme for the Dukovany (Temelin) NPP will be completed by 2008 (2016). (P.A.)

Complex monitoring of the surroundings of the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Tylova, E.

1993-01-01

Based on a Resolution of the Government of the Czech Republic, the Ministry of the Environment and the Ministry of Health of the Czech Republic shall develop a project of complex environmental pollution and contamination monitoring in the surroundings of the Dukovany nuclear power plant and shall discuss this project with municipalities there till the end of 1993. The objective of the project is to assess in a complex manner the situation in the Dukovany area with respect to all risks and their simultaneous effects, so as to ensure that the population in the area concerned is not burdened to an intolerable extent. (Z.S.)

Human factor in the operation of the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Kostiha, Frantisek; Pleskac, Frantisek

2009-01-01

The human factor, i.e. the action of man within complex technical systems, has been in the focus of the Dukovany NPP management constantly. The paper gives an overview of the plant strategy regarding human factor issues, such as training, human factor prevention methods and practices to improve the resistance of the system to human error, the use of information systems, and operational feedback from the role of the human factor and influence of the operators on the initiation, development and resulting level of severity of operational events. The method of monitoring and assessment of the quality of human performance at the Dukovany plant on an ongoing basis aimed at a constant improvement is highlighted. (orig.)

Performance testing of self-powered detector signal converters at Dukovany nuclear power plant - stage 1

International Nuclear Information System (INIS)

Erben, O.; Hajek, P.; Zerola, L.; Karsulin, M.

1990-11-01

The converters were manufactured at the Institute of Nuclear Research, Rez. Dynamic functions of the converters were tested during the start-up of reactor unit 4, Dukovany nuclear power plant, and their stability during its normal operation. The results and evaluation of the measurements show a good performance of converters. They have a low offset, good stability and the values of current are in a good agreement with the values obtained using other methods. The values of insulation resistance are in a good agreement with the values obtained manually using the method of additional resistance. These converters are planned to be used in the upgraded in-service inspection system in WWER-440 nuclear power plants. (Z.S.) 9 tabs., 22 figs., 1 ref

NPP I and C system modernizations in the Czech Republic. The NPP Dukovany example

International Nuclear Information System (INIS)

Krs, P.

1998-01-01

There are four units of WWER 440/213 type reactors under operation at Nuclear Power Plant Dukovany site in the Czech Republic. The EEZ utility has decided to include upgrade of existing Instrumentation and Control (I and C) systems as one of the most significant parts of a larger scale modernization project. The original I and C systems designed in the late 70's and early 80's with analogue equipment and relays will be subjected to an integrated modernization programme developed to replace obsolete equipment, to balance operational and maintenance costs, to improve performance and to enhance plant safety. To achieve this objective within next decade, the utility has already started preparatory phase of the overall modernization project, including analytical and planning activities related to the I and C part. (author)

Programs of plant life management at NPP Dukovany

International Nuclear Information System (INIS)

Pochman, K.; Ruscak, M.; Brumovsky, M.

2000-01-01

The program of lifetime management at NPP Dukovany is oriented towards an effective usage of facilities while keeping their safety parameters, mainly concerning nuclear safety. The goal is to achieve maximum usage of the design lifetime under the condition of fulfilling above mentioned requirements. Evaluation procedures have been prepared for selected components. These are subsequently discussed and approved by the State Office of Nuclear Safety. The databases of original data are built up regarding the evaluated components. The input data are completed as well according to the approved procedures. Generally valid approaches based on the evaluation of all possible degradation mechanisms are modified regarding the original status of particular component. The procedures are oriented towards a quantification of residual lifetime. They employ knowledge of initiation and kinetics of degradation mechanisms. They also define mechanisms, which should be quantified. For irreplaceable components or components whose replacement is difficult, the program is focused to the control of operation conditions and inspection of facilities. However, they are again based on the knowledge of degradation kinetics. If at least some parts of the facility could be replaced, better material selection regarding the operation conditions is an option. Systems included into the ageing management system will be specified in the paper. Also, the measures will be identified which serve as a check of implementation of all possible stressors and degradation mechanisms into the first screening. The examples of lifetime evaluation of some systems and components will be given in the paper: reactor pressure vessel, reactor internals, steam generator tubes and feedwater piping. Specific approaches will be shown fro different original status of components and different operation conditions. (author)

Modifications of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany based upon new version of Emergency Operating Procedures

International Nuclear Information System (INIS)

Aldorf, R.

1997-01-01

In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to reflect on Probabilistic Safety Assessment-1 basis on impact of Emergency Response Guidelines (as one particular event from the list of other modifications) on Plant Safety. Following highlights help to orient the reader in main general aspects, findings and issues of the work that currently continues on. Older results of Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany have revealed that human behaviour during accident progression scenarios represent one of the most important aspects in plant safety. Current effort of Nuclear Power Plants Dukovany (Czech Republic) and Bohunice (Slovak Republic) is focussed on development of qualitatively new symptom-based Emergency Operating Procedures called Emergency Response Guidelines Supplier - Westinghouse Energy Systems Europe, Brussels works in cooperation with teams of specialist from both Nuclear Power Plants. In the frame of 'living Probabilistic Safety Assessment-1 Nuclear Power Plant Dukovany Project' being performed by Nuclear Research Institute Rez during 1997 is planned to prove on Probabilistic Safety Assessment -1 basis an expected - positive impact of Emergency Response Guidelines on Plant Safety, Since this contract is currently still in progress, it is possible to release only preliminary conclusions and observations. Emergency Response Guidelines compare to original Emergency Operating Procedures substantially reduce uncertainty of general human behaviour during plant response to an accident process. It is possible to conclude that from the current scope Probabilistic Safety Assessment Dukovany point of view (until core damage), Emergency Response Guidelines represent adequately wide basis for mitigating any initiating event

Upgrading the standard makeup node at the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Charvat, L.; Dvorak, J.

1999-01-01

Modifications made on the normal makeup and boron control system of the Dukovany nuclear power plant in 1997 and 1998 are described. The reasons for and objectives of the upgrading are highlighted in relation to the reliability of the facility and maintenance cost of the system. (author)

Methods and tools for the validation of neutron instrumentation; methods for the detection of loose VVER-1000 reactor internals. Technical report

International Nuclear Information System (INIS)

Stulik, P.; Sipek, B.; Pecinka, L.

2004-12-01

The following topics are addressed: (1) Development, tuning and laboratory testing of the proposed DMTS distributed system; (2) Testing of selected technological equipment and software within the technology of the Temelin NPP; (3) Proposal for basic performance testing of the temperature measurement dynamics on Temelin primary circuit loops; (4) Data for the design and manufacture of 2 measuring chains for the processing of operating signals from internal reactor detectors at the Dukovany-4 reactor unit using a modified experimental AMV set and the DMTS system being developed; (5) Trial measurement with the DMTS system; (6) Evaluation of the usability of signals from the ionization chambers of the innovated instrumentation and control system within the in-service diagnosis system of the Dukovany NPP using the DMTS system being developed; and (7) Calculation of acoustic frequencies of the Temelin primary circuit by means of electromechanical analogy for loop configurations including the effects of the pressurizer and idle coolant loops. (P.A.)

ASTEC applications to VVER-440/V213 reactors

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: ivstt@nextra.sk; Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir

2014-06-01

Since the beginning of ASTEC development by IRSN and GRS the code was widely applied to VVER reactors. In this paper, at first specific features of VVER-440/V213 reactor design that are important from the modelling point of view are briefly described. Then the validation of ASTEC code with focus on its applicability to VVER reactors is briefly summarised and the results obtained with the ASTEC V2.0-rev1 version for the ISP-33 PACTEL natural circulation experiment are presented. In the next section the application of ASTEC V2.0-rev1 code in upgrade of VVER-440/V213 NPPs to cope with consequences of severe accidents is described. This upgrade includes adoption of in-vessel retention via external reactor vessel cooling and installation of large capacity passive autocatalytic recombiners. Results of analysis with focus on corium localisation and stabilisation inside reactor vessel, hydrogen control in confinement and prevention of long-term confinement pressurisation are presented.

Preparation of NPP Dukovany periodic safety review

International Nuclear Information System (INIS)

Dubsky, L.; Vymazal, P.

2004-01-01

Dukovany NPP in Czech Republic performs a periodic safety review for the second time after approximately 20 years of operation. The history of the Safety Report and its transformation into an internationally accepted form complying with IAEA standards is described. The deterministic and probabilistic assessment of the plant's safety-related design and state is applied to determine whether and to what extend the relevant protective goals are fulfilled by the existing plant design. A description of the step-by-step process is presented together with the creation of methods and criteria for PSR evaluation prepared by Nuclear Research Institute Rez

Durability of Dukovany shallow land repository engineered barriers. Appendix 7: Czech Republic

International Nuclear Information System (INIS)

Vokal, A.; Nachmilner, L.; Wasserbauer, R.; Dohnalek, J.

2001-01-01

The main aim of this project was to explore the durability of engineering barriers used at Dukovany shallow land repository as a support of safety assessments. This appendix summarises the principal results focused on durability of asphaltopropyleneconcrete (APC) hydroisolation and steel reinforced concrete construction

Activities of the diagnostic department at the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Polak, L.; Sadilek, J.; Spidla, M.; Zhanalova, J.

2001-01-01

Activities of the diagnosis department at the Dukovany nuclear power plant are described. The department staff monitors vibrations of rotary machines and vibrations and loose parts in the primary circuit, and inspects the condition of the electrical equipment by means the thermography technique. Examples of successful measurements and analyses of the condition of various machines are given. (author)

Analysis of external events - Nuclear Power Plant Dukovany

International Nuclear Information System (INIS)

Hladky, Milan

2000-01-01

PSA of external events at level 1 covers internal events, floods, fires, other external events are not included yet. Shutdown PSA takes into account internal events, floods, fires, heavy load drop, other external events are not included yet. Final safety analysis report was conducted after 10 years of operation for all Dukovany operational units. Probabilistic approach was used for analysis of aircraft drop and external man-induced events. The risk caused by man-induced events was found to be negligible and was accepted by State Office for Nuclear Safety (SONS)

Nuclear power plant V-2

International Nuclear Information System (INIS)

1998-01-01

The nuclear power plant Bohunice V -2 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - December 1976; first controlled reactor power - 7 August 1984, 2 August 1985; connection to the grid - 20 August 1984, 9 August 1985; commercial operation - 14 February 1985, 18 December 1985. This leaflet contains: NPP V-2 construction; Major technological equipment [WWER 440 V230 type reactor; Nuclear Power plant operation safety (Safety barriers; Safety systems [Active safety systems, Passive safety systems]); Centralized heat supply system; Scheme of Bohunice V-2 NPP and technical data

Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

International Nuclear Information System (INIS)

Kral, P.

1998-01-01

Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

Analysis of the effect of implemented low temperature overpressure regimes on the reactor pressure vessel resistance to damage

International Nuclear Information System (INIS)

Pistora, V.

1995-12-01

The temperature and stress fields of the Dukovany WWER-440 reactor pressure vessel (RPV) were calculated based on a two-dimensional model using the finite element method. Two pressurized thermal shock events occurred at Dukovany in 1992: the temperature in 3 loops dropped rapidly while the primary circuit was fully pressurized. The calculation revealed that the first event was intolerable with respect to the RPV resistance to brittle fracture; had the two events occurred towards the end of the RPV lifetime, both would have been intolerable. (M.D.). 6 tabs., 15 figs., 6 refs

Analysis of the ASME methodology for evaluation of erosion-corrosion defects with respect to the differences in the calculation and in materials used at the Dukovany NPP

International Nuclear Information System (INIS)

Kadecka, P.

1995-01-01

The problem of evaluation of tolerable defects and thinning of pipe walls was analyzed. In fact, a procedure for evaluation of tolerable defects is described in ASME Code Case N 480 based on the ASME ''Rules for Construction of Nuclear Power Plant Components''. The pipe systems of the Dukovany NPP, however, were constructed to different (East European) standards, and therefore caution should be exercised when applying US standards to this plant. The report demonstrates major differences between the ASME Standard and the proposed Czech standard ''A.S.I. Standards Documentation for Strength Calculations of Equipment and Piping of WWER Type Nuclear Power Plants'' developed by the Czech Association of Mechanical Engineers (A.S.I), evaluates the applicability of Code Case N 480 to the Dukovany plant, and proposes a Czech procedure for the evaluation. The basic characteristics of materials cited by ASME II and carbon steels used in the secondary circuit of the Dukovany NPP are also compared. (P.A.). 78 tabs., 2 figs., 4 refs

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

Nuclear power plant V-2

International Nuclear Information System (INIS)

1999-01-01

In this leaflet the short history of commissioning of Bohunice V-2 NPP is reviewed (beginning of construction December 1976; First controlled reactor power, Reactor Unit 1 (RU1): 7 August 1984, Reactor Unit 2 (RU2): 2 August 1985; Connection to the grid: RU1 20 August 1984, RU2 9 August 1985; Commercial operation: RU1 14 February 1985, RU2 18 December 1985. The scheme of the nuclear reactor WWER 440/V213 is depicted. The major technological equipment are described. Principles of nuclear power plant operation safety (safety barriers, active and passive safety systems, centralized heat supply system, as well as technical data of the Bohunice V-2 NPP are presented

Care of personnel on the building site of the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Kurial, P.

1984-01-01

The accommodation is described of workers on the building site of the Dukovany nuclear power plant. The quality is appraised of accommodation, catering and refreshments. There is a health care unit on site and 15 beds are reserved at the Trebic hospital for emergency cases. Trade union and youth organizations look after sports and cultural activities. (E.S.)

The evolution of the fuel cycle in the Dukovany NPP

Energy Technology Data Exchange (ETDEWEB)

Bajgl, J [Jaderna Elektrarna, Dukovany (Czech Republic)

1994-12-31

The ten-year operational experience of four WWER-440 units in Dukovany NPP with a total number of 35 cycles is outlined. The strategy of fuel reloading has been changed from out-in schemes to low-leakage patterns. The linear pin power limitation will be introduced. The main physical limits and conditions for the NPP operation are listed. The main goal is to go to a full 4-year fuel cycle in which the burnup will be about 40 Mwd/kg U. 6 tabs.

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

The development of public relations at the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Rerucha, F.; Urbasek, R.

2000-01-01

The article describes the development of mutual relations between the Dukovany NPP and the public and institutions within the region and presents evaluation of the current status and topical tasks for the benefit of a further progress in such relations. The importance is stressed of the role of the communication process, preparation and purposefulness of the forms and methods in implementing the communication policy by the NPP's Public Relations department towards individual target groups of the public. (author)

Diagnostic system for process control at NPP Dukovany load follow

International Nuclear Information System (INIS)

Rubek, J.; Petruzela, I.

1998-01-01

The NPP Dukovany is being operated in the frequency control since 1996. In last year a project for the plant load follow has been developed. One part of the project is to install a diagnostic system for process control. At present the main control loops of the plant control system are regular tested after unit refuelling only. The functionality and control system parameter adjusting is tested by certificated procedures. This state is unsuitable in view of the plan load follow operation. The relevant operational modes are based on minimisation of influence on plant component life time and on achievement of planned unit parameters. Therefore it is necessary to provide testing of main control system parts in shorter time period. Mainly at time when the unit is really in load follow operation. The paper describes the diagnostic system for process control which will be at NPP Dukovany implemented. The principal of the system will be evaluation of real and expected changes of technological variables. The system utilises thermohydraulic relation among main technological variables and relation among controlled and manipulated variables. Outputs of the system will be used to operational staff support at the plant operation. It enables: determination of control system state, estimation and check of future control system state, early indication of the deviation of process from normal conditions, check of efficiency of operational staff intervention into plant control. The system gives the plant operator new information for the plant process control. Simultaneously the coupling of new system outputs on existing signalisation is solved. (author)

Thermal performance monitoring and assessment in Dukovany nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Madron, F. [Chemplant Technology s.r.o., Hrncirska 4, 400 01 Usti nad Labem (Czech Republic); Papuga, J. [CEZ a.s., JE Dukovany, 675 50 Dukovany (Czech Republic); Pliska, J. [I and C ENERGO a.s., Prazska 684, 674 01 Trebic (Czech Republic)

2006-07-01

Competition in the European electricity market forces generators to achieve - in compliance with safety and environmental standards - efficiency of production as high as possible. This efficiency or heat rate is an important indicator of both the condition of the plant equipment and the quality of plant operation. Similar thermal performance indicators can also be calculated for components of the plant equipment such as heat exchangers. However, it is not easy to quantify these indicators with sufficient precision so that the results can be used for conduct of plant operation in near-real time and for predictive maintenance. This paper describes a present state of the system monitoring and evaluating thermal performance of the reactor units in Dukovany Nuclear Power Plant. The system provides information on actual and desirable (should-be) values of thermal performance indicators for control room operators, performance engineers and maintenance planners. The system is designed to monitor steady states and has two main functions: data validation and process simulation. Data validation is based on data reconciliation methodology and carried out with Recon software by Chemplant Technology. A detailed model of the secondary side for mass and heat balancing has been made up by means of the Recon's graphical editor; now it contains roughly 300 flows and employs data of about 200 measurements. Main advantages of the data reconciliation are: - reconciled data are consistent with the model, - reconciled data are more precise than data directly measured with consequence that the thermal power of steam generators is determined with substantially lower uncertainty than before - data reconciliation represents a solid basis for detection and identification of data corrupted by gross errors. Simulation is performed with a different analytical model of plant components configured into secondary side. The model has been developed by I and C Energo. Main purposes of simulation

Use of TL dosemeters for measuring doses of external gamma radiation in the vicinity of the Dukovany NPP

International Nuclear Information System (INIS)

Ohera, M.; Fiala, E.; Valasek, J.

1992-01-01

The results of measurement of the photon dose equivalent from external gamma radiation in the surroundings of the Dukovany nuclear power plant are given for 20 sites in southern Moravia over the 1984-1990 period, i.e., one year before starting up the plant and 6 years of its operation. CaSO 4 :Dy TL dosemeters were used, applying filtration with 0.5 mm Pb and 0.2 mm Pb + 0.6 mm Sn. Since 1989, the towns along the borders with Austria (Znojmo, Vranov, Mikulov) have been included, along with the Brno, Trebic, Zakrany, Ivancice and Jaromerice sites, into the national monitoring TL dosimetric network. The elevated doses at some sites come from the bedrock and are not caused by the operation of the plant. The effect of the Chernobyl accident on the dosemeter response is also evaluated. Comparison of the data prior to the Dukovany plant start-up (1984) and during the operation (1985-1990) give evidence that the operation of the power plant did not bring about any increase in the values monitored. (Z.S.). 3 tabs., 8 figs., 9 refs

Increasing of leak-tightness of hermetic zone and other important components in NPPs

International Nuclear Information System (INIS)

Murani, J.

2001-01-01

The performed by VUEZ resealing works in the hermetic compartments/containment of WWER 440 and WWER 1000 NPPs in Slovakia (Jaslovske Bohunice V-1 NPP, Jaslovske Bohunice V-2 NPP, Mochovce NPP), the Czech Republic (Dukovany NPP, Temelin NPP) and Hungary (Pask NPP) are presented. The leakage rate from hermetic compartments at individual NPP reactor units are shown.The VUEZ activities in field of resealing of weld joints in pools and tanks in the Jaslovske Bohunice V-1 NPP and innovation of the seal of reactor pit protective hood (kolpak) are described. The principle of operation and practical experience of the SMU-V system implemented in the Jaslovske Bohunice V-1 NPP (WWER 440 - V 230) for NPP primary circuit leakage monitoring are also given

The common project for completion of Bubbler Condenser Qualification (Bohunice, Mochovce, Dukovany and Paks NPPs)

International Nuclear Information System (INIS)

Jaroslav, H.; Pavol, B.

2003-01-01

Described is the common project for completion of bubbler condenser qualification for nuclear power plants in Bohunice, Mochovice, Dukovany and Paks. Functionality of the bubbler condenser was elaborated during the simulation of the main steam line brake, medium break and small break LOCA. On this basis the appropriate operation of bubbler condenser containment under accident conditions can be positively confirmed

Research on V and V strategy of reactor physics code of COSINE

International Nuclear Information System (INIS)

Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei

2013-01-01

Verification and validation (V and V) is very important for the software quality assurance. Reasonable and efficient V and V strategy can achieve twice the result with half the effort. Core and system integrated engine for design and analysis (COSINE) software package contains three reactor physics codes, the lattice code (LATC), the core simulator (CORE) and the kinetics code (KIND), which is called the reactor physics subsystem. The V and V strategy for the physics subsystem was researched based on the foundation of scientific software's V and V method. The module based verification method and the function based validation method were proposed, composing the physical subsystem V and V strategy of COSINE software package. (authors)

Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

International Nuclear Information System (INIS)

Rebreyend, P.; Burel, J.P.; Spoc, J.; Karasek, A.

2010-01-01

Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

The effect of nuclear facilities operation on fish populations and the Dukovany-Dalesice power complex under construction

International Nuclear Information System (INIS)

Penaz, M.

1979-01-01

The scope and the main results are described of the hydrobiological and ichthyological research into the Jihlava river in the neighbourhood of the Dukovany - Dalesice power plant complex. The effect of the power plant complex on the ecosystem of the affected river stretch is predicted, mainly on the fish community. (author)

Assessment of the integrity of WWER type reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.

1995-01-01

Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

Preparation of Long Term Operation in Dukovany NPP, Czech Republic

International Nuclear Information System (INIS)

Krivanek, R.; Sabata, M.

2012-01-01

Dukovany NPP in the south-east of the Czech Republic operates four VVER 440/213 type units. The first unit was commissioned in 1985 and the last one in 1987. The operational results of the whole NPP have been excellent and NPP permanently belongs between the first quartile of the best operated NPPs in the world in accordance with WANO factors. Large safety improvement programme have been implemented in last 15 years. The original design lifetime of main components is 30 years which means till 2015 and it is understandable that NPP is preparing for long-term operation (LTO). The paper is describing activities carried out and planned for safe and successful LTO. (author)

Construction, start-up and operation of Dukovany nuclear power plant

International Nuclear Information System (INIS)

Hornicek, Z.

1989-01-01

The Labor Safety Inspectorate have been supervising the construction of the Dukovany nuclear power plant since the construction start in 1977. It was found that in concreting the reactor building walls with concrete mixes, the regulations were not observed on the highest level for the concrete mix drop, and on the gap processing and concrete treatment when concreting was interrupted. Thus, concreting was halted until the conditions for concreting were met. Attention was focused on the protection of the hermetic casing from damage, which had very often happened. A number of shortcomings were detected in storing technology parts. The cleanliness was inspected of the facilities being assembled. The inspections also revealed shortcomings in sealed space tightness. The inspections of assembly and testing of facilities showed failures of the facilities themselves (control valves, electric motors, filter and pump sealing) and of the assembly process. Faults were also detected in electrical equipment. Only a very small part of the installation showed lifetime as specified by Decree 105/1982 Coll. laws on safety assurance in the nuclear power industry. Missing data in documentation led to delays in the start-up stages. The State Surveillance Body also inspected the results of equipment testing. Prior to physical start-up, all production facilities and buildings were inspected, labor safety was inspected for all personnel in communicating corridors, staircases, manholes and onservice and handling posts. Shortcomings were removed. The organization of assembly work was a considerable problem if staff from more organizations had to work together in the same workplace. A list of tasks is presented of State Surveillance Body in operation, maintenance, outages, repairs, and troubleshooting in a nuclear power plant. (J.B.)

Quality control and assurance in building the Dukovany nuclear power plant - check on and filing of quality certificates

International Nuclear Information System (INIS)

Sokola, J.

1986-01-01

The following documents were used for determining the range of documentation required for the construction of the Dukovany nuclear power plant: the decree valid in Czechoslovakia for all industrial structures, the respective Czechoslovak State and branch standards and several special Soviet regulations. For the central recording of all documents on the quality of deliveries and assemblies a special quality assurance unit was set up on the site of the nuclear power plant. In the system of the flow of documents on the quality of the structure of the Dukovany nuclear power plant there are 16 addressees, from outside subcontractors to on-site managerial staff, work safety inspectors, etc., to the enterprise archives and the department of scientific and technical information. A brief description is presented of the different types of documents on the quality of deliveries and assemblies, and the method of inspection of their content and completeness is described. (Z.M.)

Determination of reactor thermal power using a more accurate method

International Nuclear Information System (INIS)

Papuga, J.; Madron, F.; Pliska, J.

2005-01-01

Reactor thermal power is an important operational parameter in many respects such as nuclear safety, reactor physics or evaluation of turbine thermal performance. Thermal power of a pressurized water reactor is determined on the basis of the steam generator thermal balance. The balance can be made in several variants differing from one another by the selection of different measuring circuits whose data are used in the balancing. In principle, no one such variant gives the true value of the thermal power. Among the variant values, the one nearest to the unknown true value of reactor thermal power is probably the value calculated with the lowest uncertainty. The determination of such uncertainty is not easy and its value can make even several percent, which has significant economic consequences. This paper presents the method of data reconciliation and its application to the data of the third of Dukovany NPP. The data reconciliation method allows to exploit all the information which process data contain. It is based on the statistical adjustment of the redundant data in such a way that the adjusted data obey generally valid laws of nature (e.g. conservation laws). Mass and energy balances based on the data not yet reconciled do not obey those laws because of measurement errors. For data reconciliation in Dukovany, a detailed model of mass and energy flows describing the 3rd unit from steam generators to alternator and condenser was set up. Laws of mass and energy conservation and phase equilibrium in water-steam systems are thus fulfilled. Moreover, the user can model momentum balances in pipelines and create other equations, which are respected during calculation. The data reconciliation is done regularly for hourly averages (Authors)

Results of trial operation of new generation assemblies with improved vibration stability for WWER-440 reactors of V-230 and V-213 plants

International Nuclear Information System (INIS)

Vasilchenko, I.; Lushin, V.; Shumeev, A.

2006-01-01

FAs, in the region of FA-and-absorber rod butt joint due to arrangement of hafnium plates in the upper part of FA casing was also solved. The correctness of solutions, laid down in the design of the new generation fuel assemblies, is successfully confirmed by the results of trial operation. It is planned to apply the new generation fuel for WWER-440 reactors of V-213 type at power units of Dukovany NPP in 2005, and in 2006 at Mochovce NPP and at Bohunice NPP V-2 units 3 and 4. In 2007 it is planned to extend an application of vibration-proof working assemblies of new generation for power units of V-230 type at Kola NPP Unit 1

A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code

International Nuclear Information System (INIS)

Weinberg, D.; Frey, H.H.; Tschoeke, H.

1993-01-01

The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)

ASTEC V2 severe accident integral code main features, current V2.0 modelling status, perspectives

International Nuclear Information System (INIS)

Chatelard, P.; Reinke, N.; Arndt, S.; Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F.; Eckel, J.; Jacq, F.; Marchetto, C.; Mun, C.; Piar, L.

2014-01-01

The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence. The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core. Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective. This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants

Quality assurance of the modernized Dukovany I and C safety system software

International Nuclear Information System (INIS)

Karpeta, C.

2005-01-01

The approach to quality assurance of the software that implements the instrumentation and control functions for safety category A as per IEC 61226, which has been adopted within the 'NPP Dukovany I and C Refurbishment' project, is described. A survey of the requirements for software quality assurance of the systems that initiate protection interventions in the event of anticipated operational occurrences or accident conditions is given. The software development process applied by the system designers and manufacturers, from the software requirements specification phase to the software testing phase, is outlined. Basic information on technical audits of the software development process is also provided. (orig.)

Charting a nuclear future for the Czech Republic

Energy Technology Data Exchange (ETDEWEB)

Mitev, Lubomir [NucNet, Brussels (Belgium)

2015-08-15

After the tender process for new reactors at Temelin was cancelled last year, Czech utility and nuclear operator CEZ is starting ''a broad discussion'' with the government on when and where to build new units. NucNet interviewed Petr Zavodsky, director of nuclear power plant construction at CEZ. He joined CEZ in 1994 as an independent verification and validation specialist for safety systems software at the Temelin nuclear power station. In June 2015, the Czech Cabinet approved a national action plan for the long-term future of nuclear energy, including plans to build new nuclear units at the existing Temelin and Dukovany sites. The plan is for at least one new reactor at Dukovany and Temelin, with a probable total of four new reactors in the long term at the two locations. Priority for construction of the first reactor will be given to the Dukovany site, where the first of four reactors currently operating will probably be shut down in 2035.

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

International Nuclear Information System (INIS)

Gwenaelle Le Dantec; Fichot, F.

2005-01-01

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

Computation code TEP 1 for automated evaluation of technical and economic parameters of operation of WWER-440 nuclear power plant units

International Nuclear Information System (INIS)

Zadrazil, J.; Cvan, M.; Strimelsky, V.

1987-01-01

The TEP 1 program is used for automated evaluation of the technical and economic parameters of nuclear power plant units with WWER-440 reactors. This is an application program developed by the Research Institute for Nuclear Power Plants in Jaslovske Bohunice for the KOMPLEX-URAN 2M information system, delivered by the USSR to the V-2 nuclear power plants in Jaslovske Bohunice and in Dukovany. The TEP 1 program is written in FORTRAN IV and its operation has two parts. First the evaluation of technical and economic parameters of operation for a calculation interval of 10 mins and second, the control of the calculation procedure, follow-up on input data, determination of technical and economic parameters for a lengthy time interval, and data printout and storage. The TEP 1 program was tested at the first unit of the V-2 power plant and no serious faults appeared in the process of the evaluation of technical and economic parameters. A modification of the TEP 1 programme for the Dukovany nuclear power plant is now being tested on the first unit of the plant. (Z.M.)

Factors contribution to the reduction of occupational exposures at Dukovany NPP

International Nuclear Information System (INIS)

Petrova, K.; Urbancik, L.; Fuchsova, D.; Kulich, V.

2008-01-01

This paper describes the causes of achieving such low collective doses and the ways used for a further optimization process reducing occupational exposure at the Dukovany NPP from the beginning of operation up to now. The objective of this contribution is to explain the approaches leading to reduced occupational radiation exposure, measures which have been implemented and what is needed for the maintenance of excellent results in the future. It is possible to identify three main areas and eight main sources affecting collective dose values. There are one objective and two specific causes for low exposures: 1. Objective cause: - General arrangement; - Structural materials; - Fuel integrity. 2. Specific cause influenced by the state regulation: - Legislative support; - Operational safety culture; - Well-though tout system of radiological monitoring; - Effective radiological event feedback; - Effective education and training. 3. Specific cause uninfluenced by the state regulation: - Modified water chemistry of the primary circuit adopted by the licensee; - Licensee's system of radiation work debriefing. All the items specified above contribute to a unique concurrence of circumstances leading to the constantly low occupational exposures at the Czech nuclear power plants. The paper discusses all factors involved in these good results. (authors)

Fish cultures in the area of the nuclear power station Dukovany and their prospects

International Nuclear Information System (INIS)

Penaz, M.

1980-01-01

The results of investigating the fishery production and measurements of radiation background in the ecosystem of the Jihlava river, in the area of a nuclear power station being built near Dukovany are evaluated with regard to the prospects of fish cultures when the power station operates in full extent. Basic data were provided for calculating the burden of significant groups of the population, resulting from the ingestion of fish of local provenance. All available information shows that under normal operating conditions, the power station should not exert a negative influence on the quality of fish food and that in the area under study it would be possible to manage intensive fish cultures. (author)

Solution of the new Dukovany benchmark using the new version of the KARATE-440 code

International Nuclear Information System (INIS)

Hegyi, G.; Kereszturi, A.; Maraczy, C.

2008-01-01

A new -so called second generation - fuel type developed by Russian vendor (TVEL) for WWER-440 has been introduced. Even the outer parameters of the assembly remain unchanged, the length of the pins was increased and 6 of the 126 pins have been doped by gadolinium. The above mentioned modifications requiring more accurate calculations have necessitated the further development and validation of our KARATE code system. The paper summarizes the capabilities of the new version of the KARATE program concerning the properties of currently used and newly developed fuel types. On the basis of an international benchmark published from Dukovany NPP validation procedure has been started. (Authors)

A successful approach for the implementation of symptom-based emergency operating procedures for VVER reactors

International Nuclear Information System (INIS)

Lhoest, V.; Prior, R.; Pascal, G.

2000-01-01

The paper provides an overview of the organization, the progress and the results of the various Emergence Operating Procedure (EOP) development programs for VVER type reactors conducted by Westinghouse so far. The detailed working process is presented through the solutions to some major plant issues. The EOPs have been developed for the Temelin, Dukovany, Bohunice, Mochovce and Paks VVER nuclear power plants. The procedures are developed in working teams of experts from the utility and Westinghouse. The completion of the programs constitute an indication of the overall success of this approach. This is further reinforced by the general acceptance of the new procedures by the plant personnel, together with the good results obtained so far from procedure testing. This is also confirmed by a new PSA-level 1 analysis for Dukovany plant, which shows a significant improvement in the overall plant safety. This means a 20% reduction in the Core Damage Frequency due to the introduction of the new EOPs. The fact that some modifications have been implemented to the plants to solve design weaknesses identified in the course of this programs also constitute a positive result

Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

Energy Technology Data Exchange (ETDEWEB)

Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble.

Comparison of the Safety Critical Software V and V Requirements for the Research Reactor Instrumentation and Control System

International Nuclear Information System (INIS)

Joo, Sungmoon; Suh, Yong-Suk; Park, Cheol

2016-01-01

This study was motivated by a research reactor project where the owner of the project and the equipment vendors are from two different standards frameworks. This paper reviews two major standards frameworks - NRC-IEEE and IAEA-IEC - and the software classification schemes as a background, then discuss the V and V issue. The purpose of this paper is by no means to solve the cross-standards-framework qualification issue, but, rather, is to remind the stakeholders of research reactor projects. V and V are also essential for the approval from regulatory bodies. As standards define or recommend consolidated engineering practices, methods, or criteria, V and V activities for software qualification are not exceptional. Within a standards framework, usually, the processes for the qualification of safety-critical software are well-established such that the safety is maximized while minimizing the compromises in software quality, safety, and reliability. When, however, multiple standards frameworks are involved in a research reactor project, it is difficult for equipment vendors to implement appropriate V and V activities as there is no unified view on this cross-standards-framework qualification issue yet. There are two major standards frameworks for safety-critical software development in nuclear industry. Unfortunately different safety classifications for software and thus different requirements for qualification are in place. What makes things worse is that (i) there are ambiguities in the standards and rooms for each stakeholders’ interpretation, and (ii) there is no one-to-one mapping between the associated V and V methods and activities. These may put the stakeholders of research reactor projects in trouble

Some causes of vibrations recorded by in-service diagnostic systems in steam generators of units 1 and 2 of Dukovany nuclear power plant

International Nuclear Information System (INIS)

Sadilek, J.; Matal, O.

1989-01-01

A brief description is presented of the design of the steam generators of the first and second units of the Dukovany nuclear power plant. Attention is also given to the feed water systems and the diagnostic systems. The causes are analyzed of the irregularly occurring vibrations in the steam generators in service. It is demonstrated that the source of the vibrations transmitted to the steam generators are the valves in the feeding tract. The vibrations are induced by dynamic forces from the feed water. Reducing the water pressure at the delivery of the electric feed pumps by reducing the size of the rotor, etc., does not remove all vibrations. It is therefore recommended that valves be ins+alled with better regulating characteristics. (Z.M.). 6 figs., 1 tab., 3 refs

Physics of Fast and Intermediate Reactors. V. I. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. V. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-03-15

in all cases that of heir presentation during the Seminar. Changes have been made where it was considered that these would enhance the usefulness of these volumes as reference books. The subject grouping adopted is given below. Volume I - I. Neutron Physics: I.1. Data requirements, I.2. Cross-section measurements, I.3. Fission properties, I.4. Nuclear theory, I.5. Multi-group cross-sections; II. Integral Experiments: II.1. Critical experiments, II.2. Other integral experiments, II.3. Theoretical correlations; Volume II - III. Reactor Theory: III.1. Calculation methods, III.2. Effects of cross-section errors, III.3. Reactivity effects, III.4. Long-term effects, III.5. Reactor concept studies; Volume III - IV. Reactor Dynamics: IV.1. Kinetics, IV.2. Stability, IV.3. Doppler effect, IV.4. Safety problems; V. Physics of Specific Reactors.

Production of 48V in a nuclear reactor via secondary tritons

International Nuclear Information System (INIS)

Siri, S.; Cohen, I.M.

2009-01-01

The production of 48 V in a nuclear reactor, induced on titanium by tritons generated from the 6 Li(n, t) 4 He reaction, and eventually 7 Li(n, n't) 4 He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from 48 V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

New code for VVER-440 loading pattern design

International Nuclear Information System (INIS)

Bajgl, J.; Lehmann, M.

1999-01-01

This paper describes the main attributes of a new computer program OPTIMAL used for loading pattern design in Dukovany NPP (4 reactors VVER-440). We have been developed this program in Nuclear Research Institute Rez since 1994 on the base of special contract between Dukovany NPP and Nuclear Research Institute Rez. General information about the optimisation methodology is given in the first part. The organisation of the optimisation process is described in part 2. Construction of the optimisation functional is shown in part 3. Procedures used during one-cycle optimisation are described in part 4. (Authors)

Quality assurance and control in constructing the Dukovany nuclear power plant

International Nuclear Information System (INIS)

Lujka, T.

1986-01-01

A quality control and assurance department was established on the Dukovany nuclear power plant site as a unit independent of the production divisions of the plant. Its responsibility consists in testing activities including the development of testing techniques and cooperation in the compilation of specifications and binding technological regulations. The department consists of five sections. The concrete laying laboratory has a staff of 7 and concentrates on testing the quality of concrete mixes and their components in the central concrete production plant. The materials testing centre with a staff of 5 provides testing of steel and special structures and oversees the laying of heavy and very heavy concretes. A separate unit for the testing of surface finish of building structures is staffed with 4 people. The section of technological checks of building and assembly work with a staff of 4 is responsible for the quality of work in the main production unit zone but also of the other buildings on the site. Two people staff the unit for checking and filing the quality control documents for selected components. (Z.M.)

Two significant events in the NPP Dukovany in 1995

International Nuclear Information System (INIS)

Dusek, J.

1996-01-01

On 13 October 1995, startup tests following refueling outage were in progress at Unit 4 of the plant. As a part of the tests, with reactor at 23% power, the neutron flux monitoring system was being checked on its output signal to the reactor power control system. When output signals from the neutron flux monitoring system to reactor protection train 2 were being tested, an operations staff member assisting with the tests inadvertently returned the reactor protection system train 2 from a ''Test'' mode into an ''Operation'' mode. As an overpower signal had been simulated into the tested train before, the reactor scram signal was formed. Due to subsequent power reduction, however, the overpower signal ceased and since the I and C technical immediately managed to recover power supply to control rod drives, the fall of control rods into the core stopped. Several seconds later, an emergency reactor period signal caused the reactor scram signal to actuate again. In the same way like previously, the control rod drive power supply was recovered and the control rods were halted second time, before they reached their lower end stops

Operational safety performance indicator system at the Dukovany Nuclear Power Plant - Experience with indicator aggregation

International Nuclear Information System (INIS)

Mandula, J.

2001-01-01

The operational safety performance indicators serve as an important tool of performance monitoring and management at the Dukovany NPP. A software-supported system has been developed, which has included: data collection, central data storage, graphic output production and periodical report generation. Analyses of performance indicator trends together with evaluation in respect of annually updated target values and acceptance criteria are used for operational safety reviews forming an integral part of continual self-assessment process. This contribution has been focused on experience obtained during development of the operational safety assessment model using indicator aggregation. It summarises problems that had to be paid specific attention in the development process. Thanks to their solution, the model has become a synoptic monitor and a useful tool for operational safety assessment. (author)

Calibration of a dePangher long counter from 2 keV to 19 MeV

Energy Technology Data Exchange (ETDEWEB)

Slaughter, D R; Rueppel, D W [California Univ., Livermore (USA). Lawrence Livermore Lab.

1977-09-01

The authors have measured the sensitivity of a dePangher precision long counter (PLC) relative to /sup 7/Li(p,n), T(p,n), and T(d,n) differential cross sections over the neutron energy range 10 keV-19 MeV. Absolute sensitivity at 2 keV is also measured, using a scandium filtered beam at a reactor. Results obtained with errors in the range 5-30% are consistent with the assumption of a nearly constant sensitivity over the range 2 keV-6 MeV, but exhibit a reduction above 12 MeV.

Plant life time management for safe long term operation

International Nuclear Information System (INIS)

Burket, Danes

2010-01-01

The topics discussed include LTO (long-term operation) and licensing procedure in general and specifically for the Dukovany NPP, components of an LTO programme, the PLIM (plant life management) concept for Dukovany and Temelin, the LTO implementation project for Dukovany, LTO strategy, LTO risk study, international verification. The Conclusions include the following items: (i) Technical-economic study of Dukovany NPP LTO demonstrates technical feasibility and economic profitability of Dukovany NPP LTO with perspective for up to 60 years of operation. (ii) Safety part of Program for assurance of Dukovany NPP LTO complies with IAEA SALTO recommendations. (iii) Dukovany NPP LTO programme incorporated IAEA SALTO Peer Review Mission recommendations. (iv) LTO Implementation Project for 2009-2015 was approved with major targets to renew operational permission and prepare NPP for operation up to 60 years. (v) Preparation of Temelin NPP LTO programme has been started. (P.A.)

Fast-reactor-data testing of ENDF/B-V at ORNL

International Nuclear Information System (INIS)

Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

1982-01-01

The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

QA/QC - Practices and procedures in WWER fuel management

International Nuclear Information System (INIS)

Keselica, M.

1999-01-01

Construction time schedule and commissioning (unit by unit) of the NPP Dukovany as well as structure of electricity generation in the CEZ in 1998 are reviewed. History of QA/QC system establishment and rules (system standards) as well as organization chart of the NPP Dukovany and quality manual of reactor physics department are presented. Standards of worker's qualification and nuclear fuel inspections are discussed. Fuel reliability indicators are presented

Fast breeder reactors: Experience and trends. V. 2

International Nuclear Information System (INIS)

1986-01-01

The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

A methodology of selection of exercises for operator training on a control room simulator and its application to the data bank of exercises at the Dukovany NPP

International Nuclear Information System (INIS)

Holy, J.

2005-07-01

The report describes the preparation of methodology for the selection of scenarios to be used during operator training on a full-scope simulator. The scenarios are selected from a data bank of scenarios, which is under preparation based on feedback from the operational history and theoretical analyses. The new methodology takes into account 3 basic attributes defining the priority for use within the training programme: frequency of occurrence, safety-related significance, and difficulty. The attributes are scored and based on a joint score, the importance of inclusion of the scenario in the training programme is also scored. The methodology was applied to the data bank of scenarios for simulation of abnormal states and incidents trained on the up-to-date simulator of the Dukovany NPP, and the results of this pilot application were made available to Dukovany operator training staff as a tool for the preparation of training plans for the years to come. The results of a PSA study are used for a non-trivial selection of the scenarios

Search for eV sterile neutrinos at a nuclear reactor — the Stereo project

Science.gov (United States)

Haser, J.; Stereo Collaboration

2016-05-01

The re-analyses of the reference spectra of reactor antineutrinos together with a revised neutrino interaction cross section enlarged the absolute normalization of the predicted neutrino flux. The tension between previous reactor measurements and the new prediction is significant at 2.7 σ and is known as “reactor antineutrino anomaly”. In combination with other anomalies encountered in neutrino oscillation measurements, this observation revived speculations about the existence of a sterile neutrino in the eV mass range. Mixing of this light sterile neutrino with the active flavours would lead to a modification of the detected antineutrino flux. An oscillation pattern in energy and space could be resolved by a detector at a distance of few meters from a reactor core: the neutrino detector of the Stereo project will be located at about 10 m distance from the ILL research reactor in Grenoble, France. Lengthwise separated in six target cells filled with 2 m3 Gd-loaded liquid scintillator in total, the experiment will search for a position-dependent distortion in the energy spectrum.

The most extensive reconstruction of nuclear power plant with VVER 440/V230 reactor

International Nuclear Information System (INIS)

Ferenc, M.

2000-01-01

The nuclear power plant V-1 Bohunice consists of two VVER-440 units with V-230 reactors. Unit 1 was commissioned in 1978 and Unit 2 in 1980. Large experience and knowledge from the operation of previous units with V-230 reactors were incorporated into the V-1 design, which resulted in a higher level of safety and operational reliability of these units. The Siemens company which won an international bidding process developed these basic goals for the Gradual Upgrading into the so called Basic Engineering (BE). For the implementation of the Gradual Upgrading in line with the BE, Rekon consortium was established consisting of Siemens and VUJE. The implementation of the Gradual Upgrading is scheduled for the time period of 1996 - 2000. Siemens was responsible for the upgrading strategy - based on the approved results of the basic engineering phase and the PSAR, the engineering and realization of all I and C improvements, and also for the seismic upgrade. VUJE's responsibility covered the detailed engineering and implementation of mechanical, electrical and civil part of upgrading measures as well as overall organisation and evaluation of verification tests. The consortium awarded contracts for final planning and design, installation services and commissioning to other Slovakian subcontractors in order to ensure the largest possible local content. The gradual reconstruction of the V-1 Bohunice with V230 reactors represents a comprehensive reconstruction of safety-related systems and equipment. Following its completion, the units will be operated with a safety level accepted internationally. (author)

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

Production of {sup 48}V in a nuclear reactor via secondary tritons

Energy Technology Data Exchange (ETDEWEB)

Siri, S. [Comision Nacional de Energia Atomica, Centro Atomico Ezeiza, Gerencia de Capacitacion, Quimica Nuclear y Ciencias de la Salud, Ezeiza, Buenos Aires (Argentina); Cohen, I.M. [Univ. Tecnologica Nacional, Dept. de Ingenieria Quimica, Buenos Aires (Argentina)

2009-07-01

The production of {sup 48}V in a nuclear reactor, induced on titanium by tritons generated from the {sup 6}Li(n, t){sup 4} He reaction, and eventually {sup 7}Li(n, n't){sup 4}He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from {sup 48}V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

The physical protection systems at the Czech nuclear installations and the Y2K issue

International Nuclear Information System (INIS)

Malach, J.; Seifert, Z.; Sedlacek, J.; Bartak, L.; Polanka, J.

1999-01-01

The paper describes the joint effort of the State Office for Nuclear Safety, EZ, a.s., Nuclear Power Plant Dukovany and EBIS Ltd., as a subcontractor, to solve the problem of the Y2K particularly in connection with real time operation of the integrated physical protection system at Nuclear Power Plant Dukovany. (author)

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

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

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

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

Testing ENDF/B-V data for thermal reactors

International Nuclear Information System (INIS)

Craig, D.S.

1982-10-01

Lattice parameters have been calculated for some thermal reactor benchmark lattices using ENDF/B-V data. These lattices were TRX-1, -2; BAPL-UO 2 -1,-2,-3; BNL-ThO 2 - 233 UO 2 -H 2 0-1,-2,-3; MIT-4,-5,-6; and PNL-31,-33,-35 (infinite lattices). In addition, parameters were calculated for 3 ZEEP lattices, 3 High-Conversion U0 2 -H 2 0 lattices, and 7 BNL-Th0 2 - 233 U0 2 -D 2 0 lattices. These calculations were made using the integral transport cell code RAHAB with the resonance reaction rates obtained using the OZMA code operating in the discrete ordinate mode. This code calculates the resonance rates allowing for the interaction of all resonances. Four group reaction rates for use in method comparisons are given for several lattices. The author discusses the use of the OZMA code for these calculations, including the choice of options and the orders of the angular quadratures, and compares results obtained using the CRNL thermal scattering data with those obtained using ENDF/B data

Status of data testing of ENDF/B-V reactor dosimetry file

International Nuclear Information System (INIS)

Magurno, B.A.

1979-01-01

The ENDF/B-V Reactor Dosimetry File was released August 1979, and Phase II data testing started. The results presented here are from Brookhaven National Laboratory only, and are considered preliminary. The tests include calculated spectrum-averaged cross sections using 235 U fission spectrum (Watt), 252 Cf spontaneous fission spectrum (Watt and Maxwellian), and the Coupled Fast Reactor Measurement Facility (CFRMF) spectrum. 6 tables

Tests of cooling water pumps at Dukovany nuclear power plant

International Nuclear Information System (INIS)

Travnicek, J.

1986-01-01

Tests were performed to examine the operating conditions of the 1600 BQDV cooling pumps of the main coolant circuit of unit 1 of the Dukovany nuclear power plant. For the pumps, the performance was tested in the permissible operating range, points were measured below this range and the guaranteed operating point was verified. Pump efficiency was calculated from the measured values. The discussion of the measurement of parameters has not yet been finished because the obtained values of the amount delivered and thus of the pump efficiency were not up to expectation in all detail. It was also found that for obtaining the guaranteed flow the pump impeller had to be opened to 5deg -5.5deg instead of the declared 3deg. Also tested were pump transients, including the start of the pump, its stop, the operation and failure of one of the two pumps. In these tests, pressures were also measured at the inlet and the outlet of the inner part of the TG 11 turbine condenser. It was shown that the time course and the pressure course of the processes were acceptable. In addition to these tests, pressure losses in the condenser and the cooling water flow through the feed pump electromotor cooler wre tested for the case of a failure of one of the two pumps. (E.S.)

V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

International Nuclear Information System (INIS)

Teuchert, E.; Hansen, U.; Haas, K.A.

1980-03-01

V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

Energy Technology Data Exchange (ETDEWEB)

Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

1996-10-01

A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

International Nuclear Information System (INIS)

Stojic, M.; Pavicevic, M.

1964-01-01

This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

Status of the PSA use in the Czech regulatory process

International Nuclear Information System (INIS)

Dusek, J.

1994-01-01

A review of previous probabilistic safety assessment (PSA) activities initiated by regulatory body and preparation of the preliminary PSA study and final PSA study (released in January 1994) for the nuclear power plant Dukovany with WWER-440 type 213 reactor is described. A brief information about the NPP Temelin (with WWER-1000) PSA Study, shutdown and PSA risk monitor current activities for the NPP Dukovany, next PSA activities in 1994 and about planned PSA activities in future is attached. (author). 21 refs

Design and testing of reactors for 735 kV

Energy Technology Data Exchange (ETDEWEB)

Erb, W; Kraaij, D J

1965-11-01

The design and testing of five large, single phase shunt reactors rated either 110 or 55 MVAR, supplied for the 735 kV system of the Quebec Hydro Electric Commission which came into operation in the autumn of 1965 are described. As these reactors are permanently connected to the transmission lines, their losses must be considered as being continuously present and must be determined exactly. In addition to the use of a new bridge method, the losses were also measured calorimetrically for the purpose of comparison, the agreement between the two tests being remarkably good. The impulse tests with full wave and chopped wave are subsequently described.

The effect of microorganisms on asphaltopropylene concrete in a radioactive waste repository. Part 2

International Nuclear Information System (INIS)

Hlavackova, I.; Hlavacek, I.; Mara, M.; Wasserbauer, R.

1993-11-01

The permeability of asphaltopropylene concrete (APC) was examined after the action of aerobic bacteria and molds, and the changes in its volume, weight and swelling capacity were recorded. APC has been used as a sealing material in low level radioactive waste pits at the Dukovany NPP repository. Results of check-up sampling of microorganisms in the repository are evaluated. Sulphate reducing bacteria, which have been detected in soil near the reactor site, were isolated and their action upon asphaltopropylene (AP) was investigated. The resistance of bitumen layers containing model waste, against the action of aerobic bacteria and molds and against water was also examined. Bitumen samples containing model waste were found to absorb water at low temperatures considerably faster than unfilled bitumen. At elevated temperatures the absorption of water is appreciable, causing high weight losses of the bituminized waste layer due to degradation. The time dependences of the bitumen sample weight at 20 degC and 60 degC in distilled and cement water are given in the Appendix. The results included in the final reports ''Investigation of the effect of microorganisms on asphaltopropylene-based insulating materials employed as sealing in the secondary radioactive waste repository at the Dukovany NPP in relation to the microbial flora present. Bacteria'' and ''Investigation of the impact of biodegradation effects of aerobic and anaerobic microorganisms including molds on asphalt and asphaltopropylene in conditions of the ground repository at the Dukovany NPP'' are also given. (J.B.). 8 tabs., 33 figs

Nuclear power plant V-1

International Nuclear Information System (INIS)

1999-01-01

In this leaflet the short history of commissioning of Bohunice V-1 NPP is reviewed (beginning of construction 24 April 1972; First controlled reactor power, Reactor Unit 1 (RU1): 27 November 1978, Reactor Unit 2 (RU2): 15 March 1980; Connection to the grid: RU1 17 December 1978, RU2 26 March 1980; Commercial operation: RU1 1 April 1980, RU2 7 January 1981. The scheme of the nuclear reactor WWER 440/V230 is depicted. The major technological equipment (primary circuit, nuclear reactor, steam generators, reactor coolant pumps, primary circuit auxiliary systems, secondary circuit, turbine generators, NPP electrical equipment, and power plant control) are described. Technical data of the Bohunice V-1 NPP are presented

Research reactor core conversion guidebook. V.2: Analysis (Appendices A-F)

International Nuclear Information System (INIS)

1992-04-01

Volume 2 consists of detailed Appendices, covering safety analyses for generic 10 MW reactor, safety analysis - probabilistic methods, methods for preventing LOCA, radiological consequence analyses, examples of safety report amendments and safety specifications. Included in Volume 2 are example analyses for cores with with highly enriched uranium and low enriched uranium fuels showing differences that can be expected in the safety parameters and radiological consequences of postulated accidents. There are seven examples of licensing documents related to core conversion and two examples of methods for determining power limits for safety specifications in the document. Refs, figs, bibliographies and tabs

National report on 'stress tests', NPP Dukovany and NPP Temelin, Czech Republic. Evaluation of safety and safety margins in the light of the accident of the NPP Fukushima. Rev. 1

International Nuclear Information System (INIS)

2012-03-01

The stress tests were performed based on European Commission requirement as a response to the Fukushima-Daiichi accident. The stress tests encompassed the Dukovany and Temelin nuclear power plants and concentrated on the potential impacts of earthquakes, flooding, extreme weather conditions, loss of electrical power and loss of ultimate heat sink, and severe accident management. (P.A.)

Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

International Nuclear Information System (INIS)

Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

1981-01-01

Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700 0 C to fluence of 3 x 10 26 n/m 2 , E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700 0 C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations

Tensile properties and microstructure of helium-injected and reactor-irradiated V-20 Ti

Energy Technology Data Exchange (ETDEWEB)

Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

1981-01-01

Mechanical properties and microstructure of vanadium-20% titanium were examined following helium-injection and reactor irradiation. Helium was injected at ambient temperature to concentrations of 90 and 200 at. ppM; neutron irradiation was at 400, 575, 625, and 700/sup 0/C to fluence of 3 x 10/sup 26/ n/m/sup 2/, E > 0.1 MeV. Cavities representing negligible volume swelling were observed in all helium-injected specimens. Degradation of mechanical properties, especially loss of ductility due to helium, occurred at temperatures of 625 and 700/sup 0/C. The levels of helium produced in the fusion spectrum can be expected to alter the response of vanadium alloys from that observed in fast reactor irradiations.

Monitoring of operational reliability of safety-related I and C subsystems at the Dukovany NPP

International Nuclear Information System (INIS)

Fuchs, P.; Sagl, P.; Zlamal, P.

2007-01-01

First, the situation existing in the data base in 1999, i.e. before the monitoring and the operational reliability monitoring concept were introduced, is highlighted. The technique of data processing is described with focus on the assessment of the relevancy of the records, component failure rate monitoring, estimation of basic statistical parameters, evaluation of the feasibility of component failure (or failure latency) detection, assessment of the mean time to repair, FMEA of the basic components (relays end measuring chains) to establish spurious signals and dangerous failure ratio. The reliability assessment of the system functions is based on structural reliability calculations (common cause failures not included). The outcomes from the operational reliability monitoring are presented in the form of a representative set of data, graphic charts and results of system function reliability assessment. Prospects for upgrading the I and C operational reliability monitoring system to the benefit of NPP Dukovany operating economy (life cycle costs evaluation, spare parts planning, RCM application) are outlined. (author)

Safety of reactors built according to earlier standards (WWER 440/V230 type)

International Nuclear Information System (INIS)

Misak, J.; Rohar, S.

1995-01-01

The problems of safety of WWER-440/V-230 type reactors are discussed, and the following conclusions are made. (1) The reactors have a very good operational record. (2) The reactors have serious design shortcomings, which should be eliminated by safety upgrading. Core damage frequency should be further reduced. (3) PSA methods constitute an appropriate tool for assessment of plant vulnerability to some initiating events and malfunctions, for prioritization of upgrading measures and for tolerability of deviations from current safety standards. (4) The most important safety merits, such as a large thermal inertia and low rupture probability, should be properly taken into account in the analysis. (5) Extensive safety upgrading is feasible and can lead to a considerable risk reduction. In certain circumstances such upgrading is the least expensive option even though the total cost is much higher than the initial plant construction cost. (6) Properly upgraded, the reactor units may be operable until better power resources are available within the country. (7) The existing gap between the technological and political judgements of nuclear safety should be reduced continuously by information exchange improvements. (8) A unified approach to nuclear safety should be adopted for all nuclear reactors (not just WWERs) built to earlier standards. 5 tabs., 1 fig

Total β-activity and the content of most important artificial radionuclides in tissues of fishes from the Jihlava river basin

International Nuclear Information System (INIS)

Wohlgemuth, E.; Penaz, M.

1987-01-01

The radioactivity of body tissues of fishes that inhabit some typical waterbodies situated in the vicinity of the Dukovany nuclear power plant, namely the Jihlava river, Dalesice water reservoir and Olesna creek, was investigated during 1982-1983. The total β-activity and the content of 90 Sr and 137 Cs isotopes were studied. The values found were very low and comparable to those existing in most non-contaminated waters. The results obtained characterize the period prior to the start of operation of the first reactor of the Dukovany nuclear power plant and thus concern the initial state of the radioactive background of those aquatic ecosystems that may be influenced by the nuclear power plant operation and liquid wastes in the near future. (author). 2 tabs., 45 refs

CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

1982-01-01

Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD

International Nuclear Information System (INIS)

Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir

2014-01-01

A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

The PLTEMP V2.1 code

International Nuclear Information System (INIS)

Olson, A.P.

2003-01-01

Recent improvements to the computer code PLTEMP/ANL V2.1 are described. A new iterative, error-minimization solution technique is used to obtain the thermal distribution both within each fuel plate, and along the axial length of each coolant channel. A new, radial geometry solution is available for tube-type fuel assemblies. Software comparisons of these and other new models are described. Applications to Russian-designed IRT-type research reactors are described. (author)

Tensile properties and microstructure of helium injected and reactor irradiated V-20 Ti

International Nuclear Information System (INIS)

Tanaka, M.P.; Bloom, E.E.; Horak, J.A.

1980-01-01

The objective of this work was to determine the effect of preinjected helium followed by neutron irradiation on the mechanical properties and microstructure of V-20% Ti. These results will be used for the evaluation of the potential use of V-20% Ti in fusion reactor service

TiO2-photocatalyzed As(III) oxidation in a fixed-bed, flow-through reactor.

Science.gov (United States)

Ferguson, Megan A; Hering, Janet G

2006-07-01

Compliance with the U.S. drinking water standard for arsenic (As) of 10 microg L(-1) is required in January 2006. This will necessitate implementation of treatment technologies for As removal by thousands of water suppliers. Although a variety of such technologies is available, most require preoxidation of As(III) to As(V) for efficient performance. Previous batch studies with illuminated TiO2 slurries have demonstrated that TiO2-photocatalyzed AS(III) oxidation occurs rapidly. This study examined reaction efficiency in a flow-through, fixed-bed reactor that provides a better model for treatment in practice. Glass beads were coated with mixed P25/sol gel TiO2 and employed in an upflow reactor irradiated from above. The reactor residence time, influent As(III) concentration, number of TiO2 coatings on the beads, solution matrix, and light source were varied to characterize this reaction and determine its feasibility for water treatment. Repeated usage of the same beads in multiple experiments or extended use was found to affect effluent As(V) concentrations but not the steady-state effluent As(III) concentration, which suggests that As(III) oxidation at the TiO2 surface undergoes dynamic sorption equilibration. Catalyst poisoning was not observed either from As(V) or from competitively adsorbing anions, although the higher steady-state effluent As(III) concentrations in synthetic groundwater compared to 5 mM NaNO3 indicated that competitive sorbates in the matrix partially hinder the reaction. A reactive transport model with rate constants proportional to incident light at each bead layer fit the experimental data well despite simplifying assumptions. TiO2-photocatalyzed oxidation of As(III) was also effective under natural sunlight. Limitations to the efficiency of As(III) oxidation in the fixed-bed reactor were attributable to constraints of the reactor geometry, which could be overcome by improved design. The fixed-bed TiO2 reactor offers an environmentally

XRD study of yttria stabilized zirconia irradiated with 7.3 MeV Fe, 10 MeV I, 16 MeV Au, 200 MeV Xe and 2.2 GeV Au ions

Energy Technology Data Exchange (ETDEWEB)

Nakano, K.; Yoshizaki, H. [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan); Saitoh, Y. [Takasaki Advanced Radiation Research Institute, Japan Atomic Energy Agency, Takasaki, Gunma 370-1292 (Japan); Ishikawa, N. [Tokai Research and Development Center, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Iwase, A., E-mail: iwase@mtr.osakafu-u.ac.jp [Department of Materials Science, Osaka Prefecture University, Sakai, Osaka 599-8531 (Japan)

2016-03-01

To simulate energetic neutron irradiation effects, yttria-stabilized zirconia (YSZ) which is one of the major materials for electrical corrosion potential sensors (ECP sensors) was irradiated with heavy ions at energies ranging from 7.3 MeV to 2.2 GeV. Ion irradiation effects on the lattice structure were analyzed using the X-ray diffraction (XRD). The increase in lattice constant was induced by the ion irradiation. It was dominated by the elastic collision process and not by the electronic excitation process. The lattice disordering which was observed as a broadening of XRD peaks was also induced by the irradiation especially for 200 MeV Xe ion irradiation. The present result suggests that the expansion and/or the disordering of YSZ lattice induced by energetic neutrons may affect the durability of a joint interface between a metal housing and YSZ membrane for the usage of ECP sensors in nuclear power reactors.

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

International Nuclear Information System (INIS)

Gencheva, Rositsa V.; Stefanova, Antoaneta E.; Groudev, Pavlin P.

2005-01-01

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is 'Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)'. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed and Bleed procedure initiation. For the purpose of this, operator action with 'Reactor vessel off-gas valve - 0.032 m' opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety

Integral activation experiment of fusion reactor materials with d-Li neutrons up to 55 MeV

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Moellendorff, Ulrich von [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Wada, Masayuki [Business Automation Co., Ltd., Tokyo (Japan)

2000-03-01

An integral activation experiment of fusion reactor materials with a deuteron-lithium neutron source was performed. Since the maximum energy of neutrons produced was 55 MeV, the experiment with associated analysis was one of the first attempts for extending the energy range beyond 20 MeV. The following keywords represent the present study: d-Li neutrons, 55 MeV, dosimetry, SAND-II, spectrum adjustment, LA-150, MCNP, McDeLi, IFMIF, fusion reactor materials, integral activation experiment, low-activation, F82H, vanadium-alloy, IEAF, ALARA, and sequential charged particle reaction. (author)

Nuclear power plant V-1

International Nuclear Information System (INIS)

1998-01-01

The nuclear power plant Bohunice V -1 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - 24 April 1972; first controlled reactor power - 27 November 1978, 15 March 1980; connection to the grid - 17 December 1978, 26 March 1980; commercial operation - 1 April 1980, 7 January 1981. This leaflet contains: NPP V-1 construction; Major technological equipment (Primary circuit: Nuclear reactor [WWER 440 V230 type reactor];Steam generator; Reactor Coolant Pumps; Primary Circuit Auxiliary Systems. Secondary circuit: Turbine generators, Nuclear power plant electrical equipment; power plant control) and technical data

Pb(n,2n) cross section at 14.1 MeV

International Nuclear Information System (INIS)

Zhou Delin

1991-01-01

Pb is a potential candidate for the neutron multiplier of the fusion reactor, so its (n, 2n) cross section at 14 MeV is of importance. This work is carried out to get a new evaluation of Pb (n,2n) cross section at 14.1 MeV for checking the angle-integrated neutron emission cross section. The equal weight averaged value 2252 +- 40 Mb is adopted in the evaluation

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

CSRL-V: processed ENDF/B-V 227-neutron-group and pointwise cross-section libraries for criticality safety, reactor, and shielding studies

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.

1982-01-01

A P 3 227-neutron-group cross-section library has been processed for the subsequent generation of problem-dependent fine- or broad-group cross sections for a broad range of applications, including shipping cask calculations, general criticality safety analyses, and reactor core and shielding analyses. The energy group structure covers the range 10 -5 eV - 20 MeV, including 79 thermal groups below 3 eV. The 129-material library includes processed data for all materials in the ENDF/B-V General Purpose File, several data sets prepared from LENDL data, hydrogen with water- and polyethyelene-bound thermal kernels, deuterium with C 2 O-bound thermal kernels, carbon with a graphite thermal kernel, a special 1/V data set, and a dose factor data set. The library, which is in AMPX master format, is designated CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data). Also included in CSRL-V is a pointwise total, fission, elastic scattering, and (n,γ) cross-section library containing data sets for all ENDF/B-V resonance materials. Data in the pointwise library were processed with the infinite dilute approximation at a temperature of 296 0 K

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Nuclear and activation characteristics of materials in 14.1-MeV and 2.5-MeV neutron field

International Nuclear Information System (INIS)

Seki, Yasushi; Takeyasu, Yuuichi.

1988-11-01

The nuclear and activation characteristics of various materials and elements of interest in terms of fusion reactor design are calculated and the results are graphically shown. The elements and materials are placed in a simple geometry modelling a blanket and shield of a fusion reactor. The neutrons with 14.1-MeV and 2.5-MeV energy are generated from the region represented as D-T and D-D plasma, respectively. The following activation characteristics after neutron irradiation are shown for each material and element; 1. Time evolution of induced activity, 2. Time evolution of decay heat, 3. Delayed gamma-ray dose distribution, 4. Decay heat distribution. In addition to the above activation characteristics, nuclear characteristics during the neutron irradiation, e.g. neutron energy spectra, neutron and gamma-ray flux distribution, nuclear heating distributions, and neutron and gamma-ray dose rate are also shown. (author)

ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

International Nuclear Information System (INIS)

Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

2009-01-01

The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

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

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

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

Modernization and safety improvement project of the NPP V-2 Jaslovske Bohunice

International Nuclear Information System (INIS)

Michal, V.; Losonsky, B.; Magdolen, J.

2000-01-01

This contribution deals with the form, present state, and results of the Nuclear Power Plants Research Institute (the Slovak acronym is VUJE - Vyskumny Ustav Jadrovych Elektrarni) participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project. A short description of VUJE history, activity, and results is also presented as well as NPPs Jaslovske Bohunice characterization. VUJE was established in 1977 and deals with scientific and research needs of nuclear power plants, such as design, construction, commissioning and operation. The next fields of VUJE activity are, NPP reconstruction, NPP personnel training, radioactive waste management technology, and NPP decommissioning. The nuclear power plant, Jaslovske Bohunice, is situated approximately 15 km from the district town of Trnava in the southwestern region of the Slovak Republic. The construction of the first Czechoslovak NPP A-1 began on this site in 1957 .The construction of the double-unit NPP V-1 with WWER-440 (type V-230) reactor began in 1972. The first unit of NPP V-1 began operation in 1978 and the second in 1980. NPPs construction on the Bohunice site continued with NPP V-2, which has two units with WWER-440 (type V-213) reactors. Unit 1 and Unit 2 of NPP V-2 were commissioned in 1984 and 1985, respectively. Slovak electric utility Slovenske elektrarne (SE) is the owner/commissioner of NPP V-2. This NPP is responsible for more than 20% of the total electrical energy production of SE, making it an essential supporter of the Slovak economy. (authors)

VAPOR PHASE OXIDATION OF DIMETHYL SULFIDE WITH OZONE OVER V2O5/TIO2 CATALYST

Science.gov (United States)

Removal of volatile and odorous compounds emissions from the pulp and paper industry usually creates secondary pollution for scrubbing and adsorption processes or sulfur poising for catalytic incineration. Product studies performed in a flow reactor packed with 10 % V2O5/TiO2 cat...

Main modelling features of the ASTEC V2.1 major version

International Nuclear Information System (INIS)

Chatelard, P.; Belon, S.; Bosland, L.; Carénini, L.; Coindreau, O.; Cousin, F.; Marchetto, C.; Nowack, H.; Piar, L.; Chailan, L.

2016-01-01

Highlights: • Recent modelling improvements of the ASTEC European severe accident code are outlined. • Key new physical models now available in the ASTEC V2.1 major version are described. • ASTEC progress towards a multi-design reactor code is illustrated for BWR and PHWR. • ASTEC strong link with the on-going EC CESAM FP7 project is emphasized. • Main remaining modelling issues (on which IRSN efforts are now directing) are given. - Abstract: A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community. Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation. Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the Gen.II–III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users’ tools are being intensified. Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and

Testing the algorithms for automatic identification of errors on the measured quantities of the nuclear power plant. Verification tests

International Nuclear Information System (INIS)

Svatek, J.

1999-12-01

During the development and implementation of supporting software for the control room and emergency control centre at the Dukovany nuclear power plant it appeared necessary to validate the input quantities in order to assure operating reliability of the software tools. Therefore, the development of software for validation of the measured quantities of the plant data sources was initiated, and the software had to be debugged and verified. The report contains the proposal for and description of the verification tests for testing the algorithms of automatic identification of errors on the observed quantities of the NPP by means of homemade validation software. In particular, the algorithms treated serve the validation of the hot leg temperature at primary circuit loop no. 2 or 4 at the Dukovany-2 reactor unit using data from the URAN and VK3 information systems, recorded during 3 different days. (author)

Software V and V of PPS for Shin-Hanul Nuclear Power Plant Units 1 and 2

International Nuclear Information System (INIS)

Park, Cheollak; Kang, Dongpa; Choe, Changhui; Sohn, Sedo; Beak, Seungmin

2013-01-01

Software V and V processes determine whether the development products of a given activity conform to the requirements of that activity and whether the software satisfies its intended use and user needs. This paper introduces the software V and V activities and tasks performed during the software development life cycle performed during the software development life cycle of the Plant Protection System (PPS) for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2). The PPS generates signals to actuate Reactor Trip (RT) and Engineered Safety Features (ESF) whenever monitored processes exceed predetermined limits, and the PPS software is classified safety critical and an independent V and V is thus required according to regulations, code and standards. The software V and V efforts, sufficiently disciplined and rigorous, are quite essential to demonstrate that the software development process is of a high quality. The software V and V of PPS for SHN 1 and 2 has been accomplished successfully with systematic V and V procedures and methods established until test phase in compliance with related code and standards. In particular, the use of automated tools such as LDRA and DOORS greatly has contributed to an improvement of a software quality, and a reduction of a verification time and human errors

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

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

The V-1 NPP and V-2 NPP upgrading

International Nuclear Information System (INIS)

1998-01-01

A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

Switching Overvoltages in 60 kV reactor compensated cable grid due to resonance after disconnection

DEFF Research Database (Denmark)

Bak, Claus Leth; Baldursson, Haukur; Oumarou, Abdoul M.

2008-01-01

power could be directly connected to long cables. Switching both cable and reactor together will cause resonance to occur between the cable capacitance and the inductance of the cable during last end disconnection. Similar type of resonance condition is known to have caused switching overvoltages...... on the 400kV grid in Denmark. Therefore it is considered necessary to analyze further whether connecting a reactor directly to 60kV cable can cause switching overvoltages. A model in PSCAD was used to analyze which parameters can cause overvoltage. The switching resonance overvoltage was found to be caused...

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

2 keV filters of quasi-mono-energetic neutrons

International Nuclear Information System (INIS)

Adib, M.; Habib, N.; El-Mesiry, M.S.; Bashter, I.I.; Saleh, A.; Fathallah, M.

2013-01-01

A simulation study for the production of 2 keV filters of quasi-mono-energetic neutrons based on the deep interference minima in the 45 Sc total cross-section was carried out. A computer code QMENF-II was adapted to calculate the optimum amounts of the 45 Sc as a main filter element and additional component ones to obtain sufficient intensity at high resolution and purity of the filtered quasi-mono-energetic neutrons. The emitted neutron spectrum from nuclear reactor and from the reaction of 2.6 MeV protons on a lithium fluoride target at the accelerator beam port, are used for simulation

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

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

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

International Nuclear Information System (INIS)

Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

The BR2 materials testing reactor. Past, ongoing and under-study upgradings

Energy Technology Data Exchange (ETDEWEB)

Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

1990-05-01

The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

Reactor plant for Belene NPP completion

International Nuclear Information System (INIS)

Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

2004-01-01

Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Further study on parameterization of reactor NAA: Pt. 2

International Nuclear Information System (INIS)

Tian Weizhi; Zhang Shuxin

1989-01-01

In the last paper, Ik 0 method was proposed for fission interference corrections. Another important kind of interferences in reator NAA is due to threshold reaction induced by reactor fast neutrons. In view of the increasing importance of this kind of interferences, and difficulties encountered in using the relative comparison method, a parameterized method has been introduced. Typical channels in heavy water reflector and No.2 horizontal channel of Heavy Water Research Reactor in the Insitute of Atomic Energy have been shown to have fast neutron energy distributions (E>4 MeV) close to primary fission neutron spectrum, by using multi-threshold detectors. On this basis, Ti foil is used as an 'instant fast neutron flux monitor' in parameterized corrections for threshold reaction interferences in the long irradiations. Constant values of φ f /φ s = 0.70 ± 0.02% have been obtained for No.2 rabbit channel. This value can be directly used for threshold reaction inference correction in the short irradiations

Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-07-01

Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S n code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S 4 approximation. Reactor cell was treated as D 2 O-U-D 2 O system. Time of iteration was 185 s [sr

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

1990-09-01

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

OpenAIRE

Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

2016-01-01

To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

A comparison of the free vacancy production in α brass by fission reactor neutrons and 14.8-MeV neutrons

International Nuclear Information System (INIS)

Damask, A.C.; Van Konynenburg, R.; Borg, R.J.; Dienes, G.J.

1976-01-01

Enhancement of substitutional diffusion is observed in α brass (30 wt% Zn) by following the decrease in electrical resistivity with neutron irradiation of a thermally equilibrated alloy; the decrease arises from the increase in short-range order. It was determined by previous research that this diffusion enhancement is largely caused by the annealling of radiation-produced vacancies in excess of the thermal equilibrium concentration. Therefore, the results reported here are based upon a well-established technique. The rate of resistivity change per neutron of different energies will give the relative number of free vacancies produced per neutron. This experiment compares the effect of 14.8 MeV neutrons with neutrons from a fission reactor. The results indicate that 14.8 MeV neutrons produce 10 +- 2 times as many free vacancies as reactor neutrons when the latter are expressed in terms of those neutrons with energies greater than 0.1 MeV. (author)

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

Energy Technology Data Exchange (ETDEWEB)

Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

1962-04-15

In search for semiconductors, which can be used in high-flux reactors in order to measure flux distributions, we irradiated SiC p-n junctions in the Belgium BR-1 reactor. Two types of SiC-diodes of different origin have been irradiated. These junctions are grown in the Lely-furnace. The change in forward and reverse characteristics have been measured during and after irradiation up to temperatures of 150{sup o}C, while measurements up to a temperature of 500{sup o}C are in progress. It has been found that one type resists BR-1 neutrons up to an integrated flux of 10{sup 15} n/cm{sup 2}, while the other resists irradiation up to a flux of 10{sup 17} n/cm{sup 2}. The changes in characteristics are given as well as the result of some annealing experiments. (author) [French] En recherchant des semi-conducteurs pouvant servir a mesurer les distributions de flux dans les reacteurs a haut flux de neutrons, les auteurs ont irradie des jonctions p-n au SiC dans le reacteur belge BR-1. Deux types de diodes a SiC d'origines differentes ont ete ainsi irradies. Les jonctions en question sont preparees par etirage dans le four Lely. Les auteurs ont mesure les modifications subies par les caracteristiques I-V apres et pendant l'irradiation a des temperatures allant jusqu'a 150{sup o}C; ils poursuivent leurs mesures dans la gamme des temperatures allant de 150{sup o}C a 500{sup o}C. Us ont constate que l'un des types de diode a SiC resiste aux neutrons du reacteur BR-1 jusqu'a 10{sup 15} n/cm{sup 2}, tandis que l'autre type resiste a l'irradiation jusqu'a 10{sup 17} n/cm{sup 2}. Les auteurs indiquent les modifications subies par les caracteristiques, ainsi que le resultat de certaines experiences de recuit. (author) [Spanish] Los autores estan tratando de encontrar semiconductores con los que sea posible medir distribuciones de flujo en reactores de flujo elevado, y con este fin irradiaron uniones p-n del SiC en el reactor BR-1 de Belgica. Irradiaron dos tipos de diodos de SiC de

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

Durbin, M.E.; Montoya, G.M.

1991-06-01

This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05; Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA, Zad. 2.50.05

Energy Technology Data Exchange (ETDEWEB)

Stojic, M; Pavicevic, M

1964-07-01

This report contains the following volumes V and VI of the Project 'Independent CO{sub 2} loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels. Ovaj izvestaj sadrzi dva albuma zadatka 'Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA', Zad. 2.50.05: Album V: Predprojekat sistema dozimetrijske kontrole u nezavisnom kolu CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA i Album VI: Izvestaj o sigurnosti za nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA.

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Determination of chemical forms of 14C in liquid discharges from nuclear power plants.

Science.gov (United States)

Svetlik, I; Fejgl, M; Povinec, P P; Kořínková, T; Tomášková, L; Pospíchal, J; Kurfiřt, M; Striegler, R; Kaufmanová, M

2017-10-01

Developments of radioanalytical methods for determination of radiocarbon in wastewaters from nuclear power plants (NPP) with pressurized light water reactors, which would distinguish between the dissolved organic and inorganic forms have been carried out. After preliminary tests, the method was used to process pilot samples from wastewater outlets from the Temelín and Dukovany NPPs (Czech Republic). The results of analysis of pilot water samples collected in 2015 indicate that the instantaneous 14 C releases into the water streams would be about 7.10 -5 (Temelín) and 4.10 -6 (Dukovany) of the total quantity of the 14 C liberated into the environment. Copyright © 2017 Elsevier Ltd. All rights reserved.

REMOCIÓN DE ARSÉNICO (V ASISTIDA POR OXIDACIÓN UV SOLAR EN UN FOTO-REACTOR TUBULAR DE SECCIÓN CIRCULAR

Directory of Open Access Journals (Sweden)

Ramiro Escalera Vásquez

2010-01-01

Full Text Available Se ha construido y caracterizado un foto-reactor tubular de sección circular para su aplicación al tratamiento de aguas subterráneas contaminadas con Arsénico, As(V, utilizando las técnica de la Remoción de Arsénico por Oxidación Solar (RAOS. El concentrador solar que posee una capacidad de radiación equivalente a 2,8 soles, fue construido reciclando materiales desechados: tubos de vidrio proveniente de lámparas de Ne y tubos de desagüe sanitario de 6” (PVC, recubiertos por láminas de aluminio. Pruebas simultáneas sin agitación,realizadas aplicando la radiación UV solar a aguas sintéticas, demostraron que la remoción de As(V en el foto-reactor es más rápida queen un tubo de vidrio sólo y en una botella PET de 2 litros, logrando remociones mayores al 98% en todos los casos. Los tiempos para la aparición de los flóculos de complejo Fe-citrato fueron de 40, 50 y 90 min respectivamente, para intensidades de radiación UVA integral (290-390 nm entre 50 y 70 Wm-2. Pruebas de irradiación seguidas de agitación controlada a 30-33 s-1 de gradiente de velocidad, demostraron que el foto-reactor acelera el proceso de formación de flóculos fácilmente sedimentables al cabo de 20-30 min de agitación. Los tiempos de irradiación óptimos para el foto-reactor, el tubo y la botella son de 15, 25 y 60 min, respectivamente. Pruebas en régimen de flujo continuo en un foto-reactor de aproximadamente 1 m2 de área, con un tiempo de residencia hidráulica (igual al tiempo de irradiación de 15 min, mostraron la formación inmediata de flóculos fácilmente sedimentables cuando se agitan a 33 s-1 durante 20-30 min, lográndose una remoción del 98,36% una concentración remanente de 16,5 mgL-1 de As(V en aguas decantadas. Esto significa que se pueden tratar aproximadamente 130 Lm-2 en una jornada de 6 horas de radiación UVA de 50-70 Wm-2 de intensidad.

MLR reactor

International Nuclear Information System (INIS)

Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

1998-01-01

The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

1995-08-01

The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

Sensing Traffic Density Combining V2V and V2I Wireless Communications

Directory of Open Access Journals (Sweden)

Julio A. Sanguesa

2015-12-01

Full Text Available Wireless technologies are making the development of new applications and services in vehicular environments possible since they enable mobile communication between vehicles (V2V, as well as communication between vehicles and infrastructure nodes (V2I. Usually, V2V communications are dedicated to the transmission of small messages mainly focused on improving traffic safety. Instead, V2I communications allow users to access the Internet and benefit from higher level applications. The combination of both V2V and V2I, known as V2X communications, can increase the benefits even further, thereby making intelligent transportation systems (ITS a reality. In this paper, we introduce V2X-d, a novel architecture specially designed to estimate traffic density on the road. In particular, V2X-d exploits the combination of V2V and V2I communications. Our approach is based on the information gathered by sensors (i.e., vehicles and road side units (RSUs and the characteristics of the roadmap topology to accurately make an estimation of the instant vehicle density. The combination of both mechanisms improves the accuracy and coverage area of the data gathered, while increasing the robustness and fault tolerance of the overall approach, e.g., using the information offered by V2V communications to provide additional density information in areas where RSUs are scarce or malfunctioning. By using our collaborative sensing scheme, future ITS solutions will be able to establish adequate dissemination protocols or to apply more efficient traffic congestion reduction policies, since they will be aware of the instantaneous density of vehicles.

Sensing Traffic Density Combining V2V and V2I Wireless Communications.

Science.gov (United States)

Sanguesa, Julio A; Barrachina, Javier; Fogue, Manuel; Garrido, Piedad; Martinez, Francisco J; Cano, Juan-Carlos; Calafate, Carlos T; Manzoni, Pietro

2015-12-16

Wireless technologies are making the development of new applications and services in vehicular environments possible since they enable mobile communication between vehicles (V2V), as well as communication between vehicles and infrastructure nodes (V2I). Usually, V2V communications are dedicated to the transmission of small messages mainly focused on improving traffic safety. Instead, V2I communications allow users to access the Internet and benefit from higher level applications. The combination of both V2V and V2I, known as V2X communications, can increase the benefits even further, thereby making intelligent transportation systems (ITS) a reality. In this paper, we introduce V2X-d, a novel architecture specially designed to estimate traffic density on the road. In particular, V2X-d exploits the combination of V2V and V2I communications. Our approach is based on the information gathered by sensors (i.e., vehicles and road side units (RSUs)) and the characteristics of the roadmap topology to accurately make an estimation of the instant vehicle density. The combination of both mechanisms improves the accuracy and coverage area of the data gathered, while increasing the robustness and fault tolerance of the overall approach, e.g., using the information offered by V2V communications to provide additional density information in areas where RSUs are scarce or malfunctioning. By using our collaborative sensing scheme, future ITS solutions will be able to establish adequate dissemination protocols or to apply more efficient traffic congestion reduction policies, since they will be aware of the instantaneous density of vehicles.

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

Science.gov (United States)

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

The experience from operation of electronic personal dosimetry system at Dukovany, Temelin and Mochovce NPPs after repair of Siemens dosemeters eliminating false doses

International Nuclear Information System (INIS)

Malysak, J.; Kocvara, S.; Jurochova, B.; Zelenka, Z.; Schacherl, M.; Zrubec, M.; Kaiser, H.

2003-01-01

This presentation summarizes the operational experience of the Electronic Personal Dosimetry Systems installed at Dukovany, Temelin and Mochovce NPPs. The system consists of three basic parts: Electronic personal dosemeters (EPD); Physical layer (HW); Logical layer (SW). Number of false doses before and after correction is presented. This presentation has demonstrated the possibilities of SEOD system and the possibility of easy dose comparison between the individual NPPs after introducing this electronic dosimetry system. Basically, the results of film and electronic dosimetry systems are according to our findings nearly identical. Electronic dosemeter sensitivity to interfering electromagnetic fields is a problem which is easily re-movable. In addition, if we know this problem, these false doses in the SEOD system can be easily revealed (e.g. by investigation of histograms) and repaired

Volatility of V15Cr5Ti fusion reactor alloy

International Nuclear Information System (INIS)

Neilson, R.M. Jr.

1986-01-01

One potential hazard from the presence of activation products in fusion facilities is accidental oxidation-driven volatility of those activation products. Scoping experiments were conducted to investigate the oxidation and elemental volatility of candidate fusion reactor alloy V15Cr5Ti as a function of time, temperature, and test atmosphere. Experiments in air and in argon carrier gases containing 10 4 to 10 1 Pa (10 -1 to 10 -4 atm) oxygen were conducted to investigate the lower oxygen partial pressure limit for the formation of a low melting point (approximately 650 0 C), high volatility, oxide layer and its formation rate. Experiments to determine the elemental volatility of alloy constituents in air at temperatures of 700 0 C to greater than 1600 0 C. Some of these volatility experiments used V15Cr5Ti that was arc-remelted to incorporate small quantities (<0.1 wt. %) of Sc and Ca. Incorporation of Sc and Ca in test specimens permitted volatility measurement of radioactive constituents present only after activation of V15Cr5Ti

Early detection of power system disturbances as a condition for safe operation of the Dukovany NPP in the 'Island operation mode'

International Nuclear Information System (INIS)

Petruzela, I.

1997-01-01

The ''Frequency Plan'' worked out for the Czech Power System specifies a set of preventive measures along with the set of the frequency ranges. This problem was solved in details for the Dukovany nuclear power plant. The design changes under preparation support a reliable as well as safe operation in the island-operation mode until a full restoration of the whole power system. The design modifications are based on the following three main innovations: An incorporation of the FREA 16 frequency relay into the protection circuits; large modifications in turbine control loops; installation of software routine for the operator to support the island operation mode. The capability of the island operation has been demanded by relevant regulations both for the operating units and for that under construction. The capability can be tested under the transition to houseloads, as well as through the direct simulation of abnormal grid conditions. 1 fig

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

Conceptual designs of power tokamak-type thermonuclear reactors

International Nuclear Information System (INIS)

Shejndlin, A.E.; Nedospasov, A.V.

1978-01-01

Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

2002-07-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-03-01

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

Minutes of the 2. Meeting of the WPRS / EGRPANS / Sodium Fast Reactor Task Force (SFR)

International Nuclear Information System (INIS)

Ivanov, Evgeny; Kereszturi, Andras; Pataki, I.; Tota, A.; Vertes, P.; Kim, Taek K.; Taiwo, T.A.; Kugo, Teruhiko; Lee, Yi Kang; Messaoudi, Nadia; Michel-Sendis, Franco; ); Pascal, Vincent; Buiron, Laurent; Varaine, Frederic; Ponomarev, Alexander

2012-01-01

Five organizations (SCK/CEN, KIT, KFKI, CEA, ANL) participated in the Sodium-cooled fast reactor (SFR) Benchmark calculations and all results were collected and compiled by CEA and ANL. The compiled results of the large size cores and medium size cores were presented by V. Pascal (CEA) and T. K. Kim (ANL), respectively. Separately, A. Kereszturi presented his recently updated results. It was observed that there is wide variation in core multiplication factor, kinetics parameters, and reactivity feedback coefficients. In particular, compared to the CEA results, ANL calculated smaller k-eff, Doppler constant, but higher sodium void worth and control rod worth. The core modeling issue (heterogeneous vs. homogeneous) and solution method (diffusion vs. transport) were identified as the potential reasons of these discrepancies, including the minor impacts from the depletion chains and lumped fission product modeling. All participants agreed that additional investigation was needed to identify the reasons of these discrepancies. In addition, V. Pascal presented the informative notes of the reactivity feedback calculations methodology proposed by CEA. This document brings together the 5 presentations (slides) given at this meeting: 1 - SFR Task Force : Core behavior during transient as a function of power size and fuel nature (L. Buiron, V. Pascal, F. Varaine); 2 - Sodium Fast Reactor core Feedback and Transient response (SFRFT) Expert Group: preliminary benchmark results for large cores (L. Buiron, V. Pascal, F. Varaine); 3 - Numerical Benchmark Results for 1000 MWth Sodium-cooled Fast Reactor (T.K. Kim and T.A. Taiwo); 4 - Preliminary results of the WPRS Sodium-Cooled Fast Reactor Benchmark problems (A. Kereszturi, I. Pataki, A. Tota, P. Vertes); 5 - SFR Task Force : proposal for Feedback coefficients estimation methodology (L. Buiron, V.Pascal, F. Varaine)

Dynamical model of computation of the rhodium self-powered neutron detector current

International Nuclear Information System (INIS)

Erben, O.; Slovacek, M.; Zerola, L.

1992-01-01

A model is presented for the calculation of the rhodium self-powered neutron detector current in dependence on the neutron flux density during reactor core transients. The total signal consists of a beta emission, prompt, and gamma component and a background signal. The model has been verified by means of experimental data obtained during measurements on the LVR-15 research reactor and at the Dukovany nuclear power plant. (author) 9 figs., 21 refs

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

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

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

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

Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

International Nuclear Information System (INIS)

Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

2007-01-01

In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

Energy Technology Data Exchange (ETDEWEB)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

Neutron dosimetry in EDF experimental surveillance programme for VVER-440 nuclear power plants

International Nuclear Information System (INIS)

Brumovsky, M.; Erben, O.; Novosad, P.; Zerola, L.; Hogel, J.; Trollat, C.

2001-01-01

Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. The chains contained different number and types of containers, the omitted ones were replaced by chain elements. All of the containers were instrumented with wire neutron fluence detectors, some of the containers in the chain had spectrometric sets of neutron fluence monitors. For the absolute fluence values evaluation it was taken into account time history of the reactor power and local changes of the neutron flux along the reactor core height, correction factors due to the orientation of monitors with respect to the reactor core centre. Unfolding programs SAND-II or BASA-CF were used. The relative axial fluence distribution was obtained from the O-wire measurements. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis on the axial positions of the sample centres and fluence values in the geometric centre of the samples was calculated making use the exponential attenuation model of the incident neutron beam. Received fast neutron fluence values can be used as reference values to all VVER-440 type 213 nuclear power plant reactors. (author)

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2010-01-01

In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

Visible-Light-Responsive Photocatalysis: Ag-Doped TiO2 Catalyst Development and Reactor Design Testing

Science.gov (United States)

Coutts, Janelle L.; Hintze, Paul E.; Meier, Anne; Shah, Malay G.; Devor, Robert W.; Surma, Jan M.; Maloney, Phillip R.; Bauer, Brint M.; Mazyck, David W.

2016-01-01

In recent years, the alteration of titanium dioxide to become visible-light-responsive (VLR) has been a major focus in the field of photocatalysis. Currently, bare titanium dioxide requires ultraviolet light for activation due to its band gap energy of 3.2 eV. Hg-vapor fluorescent light sources are used in photocatalytic oxidation (PCO) reactors to provide adequate levels of ultraviolet light for catalyst activation; these mercury-containing lamps, however, hinder the use of this PCO technology in a spaceflight environment due to concerns over crew Hg exposure. VLR-TiO2 would allow for use of ambient visible solar radiation or highly efficient visible wavelength LEDs, both of which would make PCO approaches more efficient, flexible, economical, and safe. Over the past three years, Kennedy Space Center has developed a VLR Ag-doped TiO2 catalyst with a band gap of 2.72 eV and promising photocatalytic activity. Catalyst immobilization techniques, including incorporation of the catalyst into a sorbent material, were examined. Extensive modeling of a reactor test bed mimicking air duct work with throughput similar to that seen on the International Space Station was completed to determine optimal reactor design. A bench-scale reactor with the novel catalyst and high-efficiency blue LEDs was challenged with several common volatile organic compounds (VOCs) found in ISS cabin air to evaluate the system's ability to perform high-throughput trace contaminant removal. The ultimate goal for this testing was to determine if the unit would be useful in pre-heat exchanger operations to lessen condensed VOCs in recovered water thus lowering the burden of VOC removal for water purification systems.

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Phase diagrams of systems of Sr2V2O7-M2V2O7 and of Ba2V2O7-M2V2O7 (M=Ca,Cd)

International Nuclear Information System (INIS)

Fotiev, A.A.; Zhuravlev, V.D.; Zhukov, V.P.

1982-01-01

Using the methods of X-ray phase and differential thermal anlyses phase equilibria in the systems Sr 2 V 2 O 7 -M 2 V 2 O 7 and Ba 2 V 2 O 7 -M 2 V 2 O 7 , where M--Ca, Cd, are studied, their phase diagrams being built. New double pyrovanadates Mosub(0.5)Srsub(1.5)Vsub(2)Osub(7) and MBaV 2 O 7 are found [ru

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

International Nuclear Information System (INIS)

Dimov, D.

2011-01-01

The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

Main features of buildings and structures important to safety of units V1 and V2 of Bohunice NPP

International Nuclear Information System (INIS)

David, M.

1993-01-01

The program of seismic upgrading of Bohunice NPPs has been started in the year 1989 (after finishing of new seismic input). Since that time the seismic upgrading of Main building of NPP V1 has already been realized, structural as well as technological parts. Beside that the designs of seismic upgrading of other structures of NPP V1 and V2 have been completed. It has been proved that the seismic upgrading of NPPs with reactors WWER 440 is very complicated, but still possible, even in the case with high seismic intensity. It would be not possible to fulfill this complicated task without the help of IAEA Missions. The activities of IAEA experts in the program of Bohunice NPPs upgrading are appreciated very much

Measurements of the total neutron cross-sections of U and UO2 below 2 eV at different temperatures

International Nuclear Information System (INIS)

Adib, M.; Maayouf, R.M.A.; Abdel-Kawy, A.; Ashry, A.; Abbas, Y.; Abu-Zahra, A.; Hamouda, I.

1982-11-01

The total neutron cross-sections of natural uranium and its oxide are measured using two time of flight spectrometers, installed in front of two of the ET-RR-1 reactor horizontal channels, and also by a neutron diffraction spectrometer. The measurements were carried out at room temperature in the energy range from 2 eV-0.002 eV and at 210 deg. C, for neutron energies below 0.005 eV. The coherent scattering cross-section of U was deduced both from the Bragg cut-offs observed in the behaviour of the total neutron cross-section of both U and UO 2 at cold neutron energies and the neutron diffraction pattern obtained at room temperature. (author)

Evaluation of VVER-1200/V-491 reactor pressure vessel integrity during large break LOCA along with SBO using MELCOR 1.8.6

International Nuclear Information System (INIS)

Bui Thi Hoa; Tran Chi Thanh

2015-01-01

After Fukushima accident and stress test recommended by IAEA for existing reactors, higher safety requirements are enforced upon nuclear power plants during design extension and severe accident conditions. Based on those arguments, Vietnam Government requests a lot of effective safety solutions, in designs proposed for the nuclear power plants in Ninh Thuan province of Vietnam, which can prevent the accident progression toward severe accidents and mitigate severe accident consequences. One of safety requirements is related to delay time of core melt during design extension condition. Especially, if the worst case of accidents occurs, the reactor vessel integrity must be maintained at least 24 hours from the beginning of the accident. With the aim at investigation of Reactor Pressure Vessel (RPV) integrity, in this study, MELCOR 1.8.6 code is used to evaluate the integrity of RPV lower head for VVER-1200/V-491 reactor during a Large Break Loss of Coolant Accident (LBLOCA) in combination with Station Blackout (SBO) event. The study figures out several parameters related to melt down progress such as: rupture position and rupture timing, the amount of hydrogen generated. Availability of the second stage hydro-accumulators (HA2) in the VVER-1200/V-491 is assumed as an additional improvement to delay the timing of core melt as well as to maintain the vessel integrity for long-term. (author)

Damage by radiation in structural materials of BWR reactor vessels

International Nuclear Information System (INIS)

Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2002-01-01

The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

Studies of fragileness in steels of vessels of BWR reactors

International Nuclear Information System (INIS)

Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

2003-01-01

The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

CATHARE 2 V2.5 : a fully validated CATHARE version for various applications

International Nuclear Information System (INIS)

Robert, M.; Farvacque, M.; Parent, M.; Faydide, B.

2003-01-01

The CATHARE code is devoted to best estimate calculations of thermal-hydraulic transients in nuclear reactors. The code is based on a 2-fluid 6-equation model including non-condensable gases equations and additional equations for radio-chemical components transport. It allows the coolant circuits of any reactor to be represented by assembling 1D, 0D and 3D modules. The outcome of more than 20 years of joint development effort by CEA, EDF, FRAMATOME-ANP and IRSN, is the CATHARE 2 V2.5 version, to be released at the end of 2003. The development strategy of this version was to include in one single version the functionalities and the models needed for all the various applications the CATHARE partners have to deal with, in the scope of thermal-hydraulic nuclear safety analysis. Physical assessment is a highly time consuming activity for this kind of code. The 'single version' strategy results in putting together the validation efforts required for each application, thus significantly reducing the global validation effort. It also makes the maintenance process more efficient. The main objective of CATHARE 2 V2.5 is to perform safety analyses for Pressurized Water Reactors postulated accidents or other incidents, such as primary LOCAs, SGTR, loss of RHR, secondary breaks, loss of feed-water, etc. The 3D module of CATHARE has been developed in connection with the future safety methodologies based on the 3D modelling of the vessel. The main application is the large break LOCA where significant 3D effects take place, and in particular the phenomena related to the core radial temperature profile, the 3D core quenching pattern, the bypass of the accumulator water to the break and CCFL phenomena in the down-comer during the refill phase. These effects are exemplified with a PWR large break LOCA calculation using a 3D modeling of the vessel. Another major application of this version of CATHARE is to be used without modification in the SIPA French reactor simulator. This

Research reactor core conversion guidebook. V.1: Summary

International Nuclear Information System (INIS)

1992-04-01

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

Calculation of criticality of the AP600 reactor with KENO V.a code

Energy Technology Data Exchange (ETDEWEB)

Krumbein, A; Caner, M; Shapira, M [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

1996-12-01

The Westinghouse AP600 PWR has been modeled using the KENO V.a three dimensional Monte Carlo criticality program of the SCALE-PC code system. These calculations and the use of a Monte Carlo neutron transport code such as KENO will provide us with an independent check on our WIMS/CITATION calculations for the AP600 as well as for other reactors. It will also enable us to model more complicated geometries. (authors).

PSA analysis focused on Mochovce NPP safety measures evaluation from operational safety point of view

International Nuclear Information System (INIS)

Cillik, I.; Vrtik, L.

2001-01-01

Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing.(author)

Mochovce NPP safety measures evaluation from point of view of operational safety enhancement

International Nuclear Information System (INIS)

Cillik, I.; Vrtik, L.

2000-01-01

Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing. (author)

Experience from replacement and check of thermocouples during reconstruction of in-reactor temperature measurements at Bohunice V-1 units 1 and 2

International Nuclear Information System (INIS)

Slanina, M.; Stanc, J.

2001-01-01

Replacement of thermocouples in the protection tube blocks was a key phase of the reconstruction of in-reactor temperature measurements at Bohunice V-1 with regard to the success, reliability and impact on safety of unit operation. The replacement consisted of reliable and safe withdrawal of 216 old thermocouples, their disposal and installation of new thermocouples into dry channels. In the material presented, this phase of reconstruction is described in details, with focus on the evaluation of replacement quality and check activities carried out at the new installed thermocouples. (Authors)

Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

International Nuclear Information System (INIS)

Bussac, J.; Reuss, P.

1985-01-01

This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

In-service inspections of V-230 reactor

International Nuclear Information System (INIS)

Prepechal, J.

1984-01-01

It is stated that despite certain constraints the configuration of the WWER-440 is such that it allows to make in-service inspections on a fully satisfactory scale. Three factors are discussed whose existence is necessary for the implementation of in-service inspections. The program defining the scale of inspections is satisfactory with regard to the safety and reliability of reactor operation. Its further development must result in reducing time consumption and radiation burden of personnel. Regulations for the implementation and evaluation of inspections represent the weakest link in the system of in-service inspections. At present, various organizations are dealing with the said problem within international cooperation. Equipment for in-service inspections of WWER-440 reactors is relatively good. The most important knowledge is summed up gained from the ten pre-service and in-service inspections of reactors of this type made so far. (Z.M.)

The primary circuit materials properties results analysis performed on archive material used in NPP V-1 and Kola NPP Units 1 and 2

Energy Technology Data Exchange (ETDEWEB)

Kupca, L.; Beno, P. [Nuclear Power Plants Research Institute Inc., Trnava (Slovakia)

1997-04-01

A very brief summary is provided of a primary circuit piping material properties analysis. The analysis was performed for the Bohunice V-1 reactor and the Kola-1 and -2 reactors. Assessment was performed on Bohunice V-1 archive materials and primary piping material cut from the Kola units after 100,000 hours of operation. Main research program tasks included analysis of mechanical properties, corrosion stability, and microstructural properties. Analysis results are not provided.

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1988-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

Energy Technology Data Exchange (ETDEWEB)

Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I., E-mail: paul.hungler@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

2014-07-01

In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

Enhancements to the SLOWPOKE-2 nuclear research reactor at the Royal Military College of Canada

International Nuclear Information System (INIS)

Hungler, P.C.; Andrews, M.T.; Weir, R.D.; Nielson, K.S.; Chan, P.K.; Bennett, L.G.I.

2014-01-01

In 1985 a Safe Low Power C(K)ritical Experiment (SLOWPOKE) nuclear research reactor was installed at the Royal Military College of Canada (RMCC). The reactor at nominally 20 kW thermal was named SLOWPOKE-2 and the core was designed to have a total of 198 fuel pins with Low Enriched Uranium (LEU) fuel (19.89% U-235). Installation of the reactor was intended to provide an education tool for members of the Canadian Armed Forces (CAF) and an affordable neutron source for the application of neutron activation analysis (NAA) and radioisotope production. Today, the SLOWPOKE-2 at RMCC continues to be a key education tool for undergraduate and post-graduate students and successfully conducts NAA and isotope production as per its original design intent. RMCC has significantly upgraded the facility and instruments to develop capabilities such as delayed neutron and gamma counting (DNGC) and neutron imaging, including 2D thermal neutron radiography and 3D thermal neutron tomography. These unique nuclear capabilities have been applied to relevant issues in the CAF. The analog control system originally installed in 1985 has been removed and replaced in 2001 by the SLOWPOKE Integrated Reactor Control and Instrumentation System (SIRCIS) which is a digital controller. This control system continues to evolve with SIRCIS V2 currently in operation. The continual enhancement of the facility, instruments and systems at the SLOWPOKE-2 at RMCC will be discussed, including an update on RMCC's refueling plan. (author)

Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

Science.gov (United States)

Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

2018-04-05

A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

Results of charpy V-notch impact testing of structural steel specimens irradiated at ∼30 degrees C to 1 x 1016 neutrons/cm2 in a commercial reactor cavity

International Nuclear Information System (INIS)

Iskander, S.K.; Stoller, R.E.

1997-04-01

A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at ∼ 30 degrees C (∼ 85 degrees F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 10 16 neutrons/cm 2 (> 1MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was ∼ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ∼ 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications

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

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

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

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

International Nuclear Information System (INIS)

1992-01-01

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

In-pile modelling of nuclear fuel element for the MTR type reactors. Pt. 2

Energy Technology Data Exchange (ETDEWEB)

Farhadi, Kazem [AEOI, Tehran (Iran, Islamic Republic of). Radiations Application Research School

2014-06-15

In part two of the present paper, neutronic properties of the pool-type research reactor core are used to assess the similitude laws derived for out-of-pile modelling of the fuel element. The benchmark reactor used for this purpose is an IAEA 5 MW thermal pool-type research reactor currently in operation. The neutronic properties analysis are based on typical 2 200 m/sec and neutrons having 0.025 eV energy. The non-leakage capability of the system is estimated in terms of diffusion length. Also the slowing down power and the moderating ratio of the modelled methanol coolant are calculated in terms of lethargy of the diffusing medium. It is shown that the Iron which is substituted for Aluminium cladding is a relatively low absorber of neutrons but has a high neutron leakage. Methanol which replaced ordinary water as coolant is not a suitable coolant due to high neutrons absorbing substance. It is concluded that although Iron as a cladding material and methanol as a coolant meet the modelling out-of-pile criteria but are not satisfying neutronic properties. Therefore, use of them as a model clad and coolant are not suggested for research reactors. (orig.)

Compatibility of vanadium alloys with reactor-grade helium for fusion reactor applications

International Nuclear Information System (INIS)

Bell, G.E.C.; Bishop, P.S.

1993-01-01

Tests were conducted to determine the compatibility of vanadium alloys with reactor-grade helium and to define the helium gas chemistry requirements for fusion reactors, miniature tensile specimens of V-5Cr-5Ti. V-10Cr-5Ti, and V-12.5Cr-5 Ti were exposed in a once-through system to helium with 70 vppm-H 2 (measured oxygen partial pressures of 10 -12 atm) and bottle helium (measured oxygen partial pressures of -4 atm) between 500 and 700 degree C for up to 1008 h. The weight changes in the specimens were recorded. The helium-exposed specimens were tensile tested, and the effects of exposure on mechanical properties were assessed. Exposure between 500 and 700 degree C for 1008 h in He+70 vppm-H 2 resulted in complete embrittlement of all the alloys in room temperature tensile tests. The fracture mode was primarily cleavage, probably caused by a hydrogen-induced shift in the ductile to brittle transition temperature (DBTT). Weight gains increased with temperature and were largest for the V-5Cr-5Ti alloy. Specimens exposed for 531 h between 500 and 700 degree C in bottle He exhibited two distinct fracture morphologies on the fracture surfaces. Brittle cleavage around the edges of specimens gave way to ductile dimpling in the center of the specimens. The brittle region around the periphery of the specimen is most likely the highest vanadium oxide. V 2 O 5

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

International Nuclear Information System (INIS)

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

1981-05-01

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

Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

Energy Technology Data Exchange (ETDEWEB)

Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

1964-04-15

metodos de medicion de los efectos de la reactividad en APEX, HERO y AGR, y para determinar los datos relativos a la estructura fina y a la distribucion de la potencia en los conjuntos combustibles complejos. Los trabajos se concentran principalmente en el desarrollo de un metodo capaz de sustituir al 'hetrecontrol' y al 'FTD2' , para los cuerpos de reactor cuando el combustible haya alcanzado un grado de combustion considerable. La finalidad del programa experimental con la instalacion HERO es precisamente comprobar estos metodos con cuerpos complejos, inclusive con combustible que contenga plutonio. La explotacion del reactor AGR y las mediciones fisicas que se realicen con el combustible despues de la irradiacion permitiran obtener datos adicionales sobre el efecto del plutonio. (author) [Russian] Rassmatrivaetsja tshhatel'naja jeksperimental'naja i teoreticheskaja rabota, provedennaja pri konstruirovanii, stroitel'stve i vvode v jekspluataciju Uindskejlskogo usovershenstvovannogo reaktora s gazovym ohlazhdeniem i pri okazanii obshhej pomoshhi pri razrabotke usovershenstvovannogo reaktora s gazovym ohlazhdeniem (AGR) dla proizvodstva jenergii dlja grazhdanskih celej. Daetsja opisanie znachitel'nogo ob{sup e}ma rabot, kotorye byli neobhodimy dlja razrabotki prigodnyh teoreticheskih metodov rascheta: 1) raspredelenija potoka i balansa reaktivnosti v slozhnoj aktivnoj zone reaktora; 2) raspredelenija jenergii v toplive so slozhnoj geometriej i 3) vlijanija obluchenija na toplivnye cikly i raspredelenie jenergii. V kachestve vvedenija delaetsja ssylka na jeksperimental'nye dannye i teoreticheskie metody, poluchennye v rezul'tate rabot nad uranovo-magnoksovymi sistemami, i na jeksperimental'nye dannye, poluchennye v rezul'tate Sovmestnoj jeksperimental'noj programmy Britanskoj promyshlennosti (BICEP), kotorye javilis' otpravnym momentom v razrabotke teoreticheskih metodov, primenimyh k usovershenstvovannym reaktoram s gazovym ohlazhdeniem. Dlja opredelenija parametrov

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

Hydrolytic and radiolytic degradation of TBP in TBP.30% V/V-dodecane/UO2(NO3)2.HNO3.H2O systems

International Nuclear Information System (INIS)

Barreta, L.G.

1980-01-01

The hydrolytic and radiolytic degradation of TBP is investigated in systems of TBP 30% V/V-dodecane/H 2 O . HNO 3 . UO 2 (NO 3 ) 2 by gas chromatographic determination of HDBP. No direct relation between the concentration of HDBP formed and the quantity of HNO 3 extracted by the organic phase is observed in the studies of hydrolysis of TBP. The HDBP concentration is seen to increase non-linearly with the concentration of HNO 3 extracted by the organic phase. Radiolytic studies show that for doses greater than 1 Wh/l, the concentration of HDBP formed increases with the dose absorbed by the system. Whith doses smaller than 1 Wh/l and acid concentration greater than 2 M, two distinct patterns of behavior are observed. The concentration of HDBP as a function of the radiation dose absorbed by the system presents a minimum for uranyl nitrate concentrations smaller than 0.9 M; for uranyl nitrate concentrations greater than 1.3 M the concentration of radiolytic HDBP cannot be calculated because the concentration of the hydrolytic HDBP determined is greater than the sum of the experimental concentrations of hydrolytic and radiolytic HDBP. It is known that the dose absorbed by the process solutions during the reprocessing of light water reactor fuel elements is smaller than one Wh/l. Thus, dose rates between zero and one Wh/l should be studied for this system. (Author) [pt

Safety problems of nuclear power plants with reactors of new generation; Voprosy bezopasnosti v proehktakh AEhS novogo pokoleniya s WWER

Energy Technology Data Exchange (ETDEWEB)

Fedorov, V; Rogov, M; Biryukov, G [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

1996-12-31

Modernization schemes for safety enhancement of WWER-1000 reactors are proposed. In the case of WWER-1000/V-392 it is based on introduction of additional safety systems and overall design improvement. For WWER-1100 (1000-1100 MW) the safety is enhanced by passive systems built from two-stage heat exchangers. For WWER-500/600 the use of passive safety system is extended to emergency cooling of the active zone and removal of the residual heat emissions from the reactor. The technical characteristics of the three reactors are compared. 3 figs., 1 tab.

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

ZOCO V - a computer code for the calculation of time-dependent spatial pressure distribution in reactor containments

International Nuclear Information System (INIS)

Mansfeld, G.; Schally, P.

1978-06-01

ZOCO V is a computer code which can calculate the time- and space- dependent pressure distribution in containments of water-cooled nuclear power reactors (both full pressure containments and pressure suppression systems) following a loss-of-coolant accident, caused by the rupture of a main coolant or steam pipe

Microstructure alterations in the base material, heat affected zone and weld metal of a 440-VVER-reactor pressure vessel caused by high fluence irradiation during long term operation: material: 15 Ch2MFA {approx} 0, 15 C-2,5 Cr-0, 7Mo-0,3 V; Veraenderungen der Mikrostruktur in Grundwerkstoff, WEZ und Schweissgut eines 440-VVER-Reaktordruckbehaelters, verursacht durch Neutronenbestrahlung im langzeitigen Betrieb; Werkstoff: 15 Ch2MFA {approx} 0,15 C-2,5 Cr-0, 7Mo-0,3 V

Energy Technology Data Exchange (ETDEWEB)

Maussner, G; Scharf, L; Langer, R [Siemens AG Energieerzeugung KWU, Erlangen (Germany); Gurovich, B [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

1998-11-01

Within the scope of the Tacis `91/1.1 project of the European Community, ``Reactor Vessel Embrittlement``, specimens were taken from the heavily irradiated circumferential welds of a VVER pressure vessel. The cumulated fast neutron fluence in the specimens amounts to up to 6.5 x 10{sup 19} cm{sup -}2 (E > 0.5 MeV). For the multi-laboratory, coordinated study, the specimens were cutted for mechanical testing as well as analytical, microstructural and microanalytical examinations in the base metal, HAZ and weld metal with respect to the effects of reactor operatio and post-irradiation annealing as well as thermal treatment (475 C, 560 C). The analytical transmission electron microscopy (200 kV) revealed the alterations found in the mechanical properties to be due to the formation of black dots and irradiation-induced segregations and accumulations of copper and carbides. These effects, caused by operation, (neutron radiation, temperature), are much more significant in the HAZ than in the base metal. (orig./CB) [Deutsch] Im Rahmen des von der Europaeischen Union beauftragten Tacis `91/1.1 Programms `Reactor Vessel Embrittlement` wurden Bohrkerne aus dem hochbestrahlten Rundnahtbereich eines VVER-Reaktordruckbehaelters entnommen. Die kumulierte schnelle Neutronenfluenz in diesen Proben betraegt bis zu 6,5 x 10{sup 19} cm{sup -2} (E>0,5 MeV). In einer gemeinschaftlichen Untersuchung wurden mechanisch-technologische, chemische sowie mirkostrukturelle Untersuchungen an Grundwerkstoff-, WEZ- und Schweissgutproben im vergleichbaren Ausgangs-, bestrahlten und anschliessend waermebehandelten (475 C, 560 C) Werkstoffzustand durchgefuehrt. Die analytische Durchstrahlelektronenmikroskopie (200 kV) laesst als Ursache fuer die festgestellten Veraenderungen der mechanischen Eigenschaften die Bildung von Versetzungsringen (black dots) sowie von bestrahlungsinduzierten Ausscheidungen und Anreicherungen von Kupfer in den Karbiden erkennen. Diese Effekte, als Folge der betrieblichen

A search for muon neutrino to electron neutrino oscillations at Δm² > 0.1 eV²

Energy Technology Data Exchange (ETDEWEB)

Patterson, Ryan Benton [Princeton Univ., NJ (United States)

2007-11-01

The evidence is compelling that neutrinos undergo flavor change as they propagate. In recent years, experiments have observed this phenomenon of neutrino oscillations using disparate neutrino sources: the sun, fission reactors, accelerators, and secondary cosmic rays. The standard model of particle physics needs only simple extensions - neutrino masses and mixing - to accommodate all neutrino oscillation results to date, save one. The 3.8σ-significant $\\bar{v}$_e excess reported by the LSND collaboration is consistent with $\\bar{v}$_μ →$\\bar{v}$_e oscillations with a mass-squared splitting of Δm² ~ 1 eV². This signal, which has not been independently verified, is inconsistent with other oscillation evidence unless more daring standard model extensions (e.g. sterile neutrinos) are considered.

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

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

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

A SAS2H/KENO-V methodology for 3D fuel burnup analysis

International Nuclear Information System (INIS)

Milosevic, M.; Greenspan, E.; Vujic, J.

2002-01-01

An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)

Control of WWER-440 nuclear reactor

International Nuclear Information System (INIS)

Wagner, K.; Drab, F.; Grof, V.

1978-01-01

The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

The bent crystal diffraction spectrometer at the BR2 reactor in Mol

Science.gov (United States)

Kaerts, E.; Jacobs, L.; Vandenput, G.; Van Assche, P. H. M.

1988-05-01

The DuMond-type bent crystal diffraction spectrometer installed at the BR2 reactor in Mol is presented. The spectrometer is mainly designed to study nuclear γ-transitions following thermal neutron capture. It covers the energy interval 25 ≦ Eγ ≦ 1500 keV. Instead of the traditionally used quartz crystals, a highly perfect silicium crystal is chosen as analysing crystal. Diffraction occurs from the (220) plane. The "quasi-mosaic" width, introduced by bending the crystal, is as small as 0.2″. The integrated reflecting power R of the bent crystal stays constant up to 1.5 MeV in first, 680 keV in second and 300 keV in third diffraction order. For higher photon energies, only an E-1 energy dependence is observed in second and third diffraction order. Consequently, besides improving the energy resolution, the use of these silicium crystals substantially increases the spectrometer efficiency and extends the high energy limit of bent crystal diffraction spectrometers. The diffraction angles are measured with a symmetrical interferometer system which covers an angular range of -6° to +6° with a precision of about 0.01″. Minimum diffraction line widths of 0.9″ have been measured, corresponding to an energy resolution ΔE = 1.35 × 10 -6E2n-1 keV -1. The dominant contribution to the observed line widths arises from the finite extent of the source.

Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

2008-01-15

This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

Simulation of VVER MCCI reactor test case with ASTEC V2/MEDICIS computer code

International Nuclear Information System (INIS)

Stefanova, A.; Grudev, P.; Gencheva, R.

2011-01-01

This paper presents an application of the ASTEC v2, module MEDICIS for simulation of VVER Molten core concrete interaction test (MCCI) case without water injection. The main purpose of performed calculation is verification and improvement of module MEDICIS/ASTECv2 for better simulation of core concrete interaction processes. The VVER-1000 reference nuclear power plant was chosen as SARNET2 benchmark MCCI test-case. The initial conditions for MCCI test are taken after SBO scenario calculated with ASTEC version 1.3R2 by INRNE. (authors)

VERONA V6.22 – An enhanced reactor analysis tool applied for continuous core parameter monitoring at Paks NPP

Energy Technology Data Exchange (ETDEWEB)

Végh, J., E-mail: janos.vegh@ec.europa.eu [Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Pós, I., E-mail: pos@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Horváth, Cs., E-mail: csaba.horvath@energia.mta.hu [Centre for Energy Research, Hungarian Academy of Sciences, H-1525 Budapest 114, P.O. Box 49 (Hungary); Kálya, Z., E-mail: kalyaz@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Parkó, T., E-mail: parkot@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Ignits, M., E-mail: ignits@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary)

2015-10-15

Between 2003 and 2007 the Hungarian Paks NPP performed a large modernization project to upgrade its VERONA core monitoring system. The modernization work resulted in a state-of-the-art system that was able to support the reactor thermal power increase to 108% by more accurate and more frequent core analysis. Details of the new system are given in Végh et al. (2008), the most important improvements were as follows: complete replacement of the hardware and the local area network; application of a new operating system and porting a large fraction of the original application software to the new environment; implementation of a new human-system interface; and last but not least, introduction of new reactor physics calculations. Basic novelty of the modernized core analysis was the introduction of an on-line core-follow module based on the standard Paks NPP core design code HELIOS/C-PORCA. New calculations also provided much finer spatial resolution, both in terms of axial node numbers and within the fuel assemblies. The new system was able to calculate the fuel applied during the first phase of power increase accurately, but it was not tailored to determine the effects of burnable absorbers as gadolinium. However, in the second phase of the power increase process the application of fuel assemblies containing three fuel rods with gadolinium content was intended (in order to optimize fuel economy), therefore off-line and on-line VERONA reactor physics models had to be further modified, to be able to handle the new fuel according to the accuracy requirements. In the present paper first a brief overview of the system version (V6.0) commissioned after the first modernization step is outlined; then details of the modified off-line and on-line reactor physics calculations are described. Validation results for new modules are treated extensively, in order to illustrate the extent and complexity of the V&V procedure associated with the development and licensing of the new

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Study on neutron capture cross sections using the filtered neutron beams of 55 keV and 144 keV at the Dalat reactor and related applications

International Nuclear Information System (INIS)

Vuong Huu Tan; Nguyen Canh Hai; Pham Ngoc Son; Tran Tuan Anh

2007-01-01

In this fundamental research project on nuclear physics in period of 2006, the neutron capture cross sections for the reactions of 139 La (n,γ) 140 La, 152 Sm (n,γ) 153 Sm, 191 Ir (n,γ) 192 Ir and 193 Ir (n,γ) 194 Ir have been measured at 55 keV and 144 keV by the activation method using the filtered neutron beams of the Dalat nuclear research reactor. The cross sections were determined relative to the standard capture cross sections of 197 Au. The samples and standard were prepaid from high purity (99.99%) foil of Au and natural oxide powders of La 2 O 3 , Sm 2 O 3 and IrO 2 . A high efficient HPGe detector (58%) was used to detect the gamma rays, emitted from the activated samples. The absolute efficiency curve of the detector has been precisely calibrated thanks to a set of standard radioisotope sources and a multi-nuclide standard solution, supported by IAEA. The present results were compared with the previous measurements from EXFOR-2003, and the evaluated values of JENDL 3.3 and ENDF/B-6.8. (author)

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

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

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

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

EURATOM's Programme of Participation in Power Reactor Construction; Le programme de participation d'Euratom aux reacteurs de puissance; Programma uchastiya v razrabotke ehnergeticheskikh reaktorov Evratoma; El programa de participacion de la Euratom en la construccion y explotacion de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Ramadier, R. C.; Parker, E. [Communaute Europoenne de l' Energie Atomique, Bruxelles (Belgium)

1963-10-15

One of the means used by the Commission of EURATOM to promote the development of a European nuclear industry is a programme of ''Community participation'', under which the Commission will participate in power reactor construction up to a total expenditure of 32 million European Monetary Agreement units of account. The return for this will be the acquisition of information on the design, construction, start-up and operation of such reactors. So far, proposals from three companies have resulted in the signing of contracts. These companies are: (a) The Societa Elettronucleare Nazionale (SENN), which is constructing a station of 150 MW(e) net in Italy, equipped with a double-cycle boiling-water reactor; (b) The Societa Italiana Meridionale Energia Atomica (SIMEA), which has undertaken to construct a station f 200 MW(e) net in Italy, equipped with a natural uranium-graphite-CO{sub 2} reactor; (c) The Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA), which has undertaken to construct, on the French-Belgian border, a station which will be equipped with a pressurized-water reactor and whose output will reach, and probably exceed, 242 MW(e) net. Further, the Commission has been requested by the Rheinisch-Westfalisches Elektrizitatswerk - Bayernwerke (RWE-BW) group and the N.V. Samenwerkende Electriciteits-Productiebedrijve to take part in the construction o f two other power reactors - the first a 237 MW(e) double-cycle boiling-water reactor, and the second a 50 MW(e) single-cycle, natural-circulation boiling-water reactor. Community participation can take various forms, one of them being the sharing of any deficit that might result from the production of electricity by the stations during their first years of operation. The effect of EURATOM's participation has been to encourage the construction of some of these nuclear power stations. Moreover, it has resulted in the gathering of extremely useful information and w ill continue to do so in the years to come

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

Science.gov (United States)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

''Sleeping reactor'' irradiations: Shutdown reactor determination of short-lived activation products

International Nuclear Information System (INIS)

Jerde, E.A.; Glasgow, D.C.

1998-01-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux (φ) of ∼ 4 x 10 14 n/cm 2 · s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of ∼ 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about ± 0.5 s) make irradiations of 9 Be(γ,n) 8 Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to ∼ 1 x 10 10 n/cm 2 · s within 1 h. By the time the fuel elements are removed, the flux has dropped to ∼ 6 x 10 8 . Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant

TEMPERA-V2. A computer program for sensitivity analysis in a cooling channel of nuclear reactors-V.1

International Nuclear Information System (INIS)

Andrade Lima, F.R. de; Alvim, A.C.M.

1986-09-01

The sensitivity of linear responses associated with physical quantities governed by non linear differential systems is studied, using perturbation theory. The equivalence and formal differences of differential and GPT formalisms are shown. Both of them and also the matricial formalism are used for sensitivity calculations of transient problems in a typical PWR coolant channel. The results obtained are encouraging with respect to the potentiality of the method for thermal-hydraulic calculations normally performed for reactor design and safety analysis. (Author) [pt

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

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

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

Hydrogen production with effluent from an ethanol–H2-coproducing fermentation reactor using a single-chamber microbial electrolysis cell

KAUST Repository

Lu, Lu

2009-06-01

Hydrogen can be produced by bacterial fermentation of sugars, but substrate conversion to hydrogen is incomplete. Using a single-chamber microbial electrolysis cell (MEC), we show that additional hydrogen can be produced from the effluent of an ethanol-type dark-fermentation reactor. An overall hydrogen recovery of 83 ± 4% was obtained using a buffered effluent (pH 6.7-7.0), with a hydrogen production rate of 1.41 ± 0.08 m3 H2/m3 reactor/d, at an applied voltage of Eap = 0.6 V. When the MEC was combined with the fermentation system, the overall hydrogen recovery was 96%, with a production rate of 2.11 m3 H2/m3/d, corresponding to an electrical energy efficiency of 287%. High cathodic hydrogen recoveries (70 ± 5% to 94 ± 4%) were obtained at applied voltages of 0.5-0.8 V due to shorter cycle times, and repression of methanogen growth through exposure of the cathode to air after each cycle. Addition of a buffer to the fermentation effluent was critical to MEC performance as there was little hydrogen production using unbuffered effluent (0.0372 m3 H2/m3/d at Eap = 0.6 V, pH 4.5-4.6). These results demonstrate that hydrogen yields from fermentation can be substantially increased by using MECs. © 2009 Elsevier B.V. All rights reserved.

Distance matters. Assessing socioeconomic impacts of the Dukovany nuclear power plant in the Czech Republic: Local perceptions and statistical evidence

Directory of Open Access Journals (Sweden)

Frantál Bohumil

2016-03-01

Full Text Available The effect of geographical distance on the extent of socioeconomic impacts of the Dukovany nuclear power plant in the Czech Republic is assessed by combining two different research approaches. First, we survey how people living in municipalities in the vicinity of the power plant perceive impacts on their personal quality of life. Second, we explore the effects of the power plant on regional development by analysing long-term statistical data about the unemployment rate, the share of workers in the energy sector and overall job opportunities in the respective municipalities. The results indicate that the power plant has had significant positive impacts on surrounding communities both as perceived by residents and as evidenced by the statistical data. The level of impacts is, however, significantly influenced by the spatial and social distances of communities and individuals from the power plant. The perception of positive impacts correlates with geographical proximity to the power plant, while the hypothetical distance where positive effects on the quality of life are no longer perceived was estimated at about 15 km. Positive effects are also more likely to be reported by highly educated, young and middle-aged and economically active persons, whose work is connected to the power plant.

X-ray digital industrial radiography (DIR) for local liquid velocity (V(LL)) measurement in trickle bed reactors (TBRs): validation of the technique.

Science.gov (United States)

Mohd Salleh, Khairul Anuar; Rahman, Mohd Fitri Abdul; Lee, Hyoung Koo; Al Dahhan, Muthanna H

2014-06-01

Local liquid velocity measurements in Trickle Bed Reactors (TBRs) are one of the essential components in its hydrodynamic studies. These measurements are used to effectively determine a reactor's operating condition. This study was conducted to validate a newly developed technique that combines Digital Industrial Radiography (DIR) with Particle Tracking Velocimetry (PTV) to measure the Local Liquid Velocity (V(LL)) inside TBRs. Three millimeter-sized Expanded Polystyrene (EPS) beads were used as packing material. Three validation procedures were designed to test the newly developed technique. All procedures and statistical approaches provided strong evidence that the technique can be used to measure the V(LL) within TBRs.

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

Preliminary design of a tandem mirror reactor

International Nuclear Information System (INIS)

Strohmayer, J.N.

1984-04-01

The purpose of this thesis is to examine the TARA mirror experiment as a possible tandem mirror reactor configuration. This is a preliminary study to size the coil structure based on using the smallest end cell axial length that physics and engineering allow, zeroing the central cell parallel currents and having interchange stability. The input powers are estimated for the final reactor design so a Q value may be estimated. The Q value is defined as the fusion power divided by the total injected power absorbed by the plasma. A computer study was performed on the effect of the transition size, the transition vertical spacing and transition current. These parameters affect the central cell parallel currents, the recircularization of the flux tube and the ratio of central cell beta to anchor beta needed for marginal stability. Two designs were identified. The first uses 100 keV and 13 keV neutral beams to pump the ions that trap in the thermal barrier. The Q value of this reactor is 11.3. The second reactor uses a pump beam at 40 keV. This energy is chosen because there is a resonance for the charge exchange cross section between D 0 and He 2+ at this energy, thus the alpha ash will be pumped along with the deuterium and tritium. The Q value of this reactor is 11.6

The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

International Nuclear Information System (INIS)

Townes, B.M.; Hilborn, J.W.

1985-06-01

A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

Thermal energy and bootstrap current in fusion reactor plasmas

International Nuclear Information System (INIS)

Becker, G.

1993-01-01

For DT fusion reactors with prescribed alpha particle heating power P α , plasma volume V and burn temperature i > ∼ 10 keV specific relations for the thermal energy content, bootstrap current, central plasma pressure and other quantities are derived. It is shown that imposing P α and V makes these relations independent of the magnitudes of the density and temperature, i.e. they only depend on P α , V and shape factors or profile parameters. For model density and temperature profiles analytic expressions for these shape factors and for the factor C bs in the bootstrap current formula I bs ∼ C bs (a/R) 1/2 β p I p are given. In the design of next-step devices and fusion reactors, the fusion power is a fixed quantity. Prescription of the alpha particle heating power and plasma volume results in specific relations which can be helpful for interpreting computer simulations and for the design of fusion reactors. (author) 5 refs

Operation of the SLOWPOKE-2 reactor in Jamaica

Energy Technology Data Exchange (ETDEWEB)

Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

2001-07-01

Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

arXiv Performance of a full scale prototype detector at the BR2 reactor for the SoLid experiment

CERN Document Server

Abreu, Y.; Arnold, L.; Ban, G.; Beaumont, W.; Bongrand, M.; Boursette, D.; Castle, B.C.; Clark, K.; Coupé, B.; Cussans, D.; De Roeck, A.; D'Hondt, J.; Durand, D.; Fallot, M.; Ghys, L.; Giot, L.; Guillon, B.; Ihantola, S.; Janssen, X.; Kalcheva, S.; Kalousis, L.N.; Koonen, E.; Labare, M.; Lehaut, G.; Manzanillas, L.; Mermans, J.; Michiels, I.; Moortgat, C.; Newbold, D.; Park, J.; Pestel, V.; Petridis, K.; Piñera, I.; Pommery, G.; Popescu, L.; Pronost, G.; Rademacker, J.; Ryckbosch, D.; Ryder, N.; Saunders, D.; Schune, M.-H.; Simard, L.; Vacheret, A.; Van Dyck, S.; Van Mulders, P.; van Remortel, N.; Vercaemer, S.; Verstraeten, M.; Weber, A.; Yermia, F.

2018-05-03

The SoLid collaboration has developed a new detector technology to detect electron anti-neutrinos at close proximity to the Belgian BR2 reactor at surface level. A 288 kg prototype detector was deployed in 2015 and collected data during the operational period of the reactor and during reactor shut-down. Dedicated calibration campaigns were also performed with gamma and neutron sources. This paper describes the construction of the prototype detector with a high control on its proton content and the stability of its operation over a period of several months after deployment at the BR2 reactor site. All detector cells provide sufficient light yields to achieve a target energy resolution of better than 20%/√E(MeV). The capability of the detector to track muons is exploited to equalize the light response of a large number of channels to a precision of 3% and to demonstrate the stability of the energy scale over time. Particle identification based on pulse-shape discrimination is demonstrated with calibration so...

Irradiation techniques at BR2 reactor

International Nuclear Information System (INIS)

Hebel, W.

1978-01-01

Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

International Nuclear Information System (INIS)

Hermansky, P.; Krajcovic, M.

2012-01-01

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

The Jules Horowitz Reactor project, a driver for revival of the research reactor community

International Nuclear Information System (INIS)

Pere, P.; Cavailler, C.; Pascal, C.

2010-01-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

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

International Nuclear Information System (INIS)

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

2016-01-01

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

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

Activities in the Czech Republic for reactor pressure components lifetime management

International Nuclear Information System (INIS)

Brumovsky, M.

1994-01-01

The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

Current and future possibilities of V2V and I2V technologies: an analysis directed toward Augmented Reality systems

Science.gov (United States)

Betancur, J. A.; Osorio-Gómez, Gilberto; Arnedo, Aida; Yarce Botero, Andrés.

2014-06-01

Nowadays, it is very important to explore the qualitative characteristics of autonomous mobility systems in automobiles, especially disruptive technology like Vehicle to Vehicle (V2V) and Infrastructure to Vehicle (I2V), in order to comprehend how the next generation of automobiles will be developed. In this sense, this research covers a general review about active safety in automobiles where V2V and I2V systems have been implemented; identifying the more realistic possibilities related to V2V and I2V technology and analyzing the current applications, some systems in development process and some future conceptual proposals. Mainly, it is notorious the potential development of mixing V2V and I2V systems pointing to increase the driver's attention; therefore, a configuration between these two technologies and some augmented reality system for automobiles (Head-Up Display and Head-Down Display) is proposed. There is a huge potential of implementation for this kind of configuration once the normative and the roadmap for its development can be widely established.

Refurbishment programme for the BR2-reactor

Energy Technology Data Exchange (ETDEWEB)

Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

1992-07-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishment programme for the BR2-reactor

International Nuclear Information System (INIS)

Koonen, E.

1992-01-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Operational report, basics of the preliminary project for reconstruction of the RA reactor ventilation system

International Nuclear Information System (INIS)

Martinc, R.

1982-01-01

This report presents the summary of results needed for designing the reconstruction of the RA reactor ventilation system. The final complete report will published later, including results of possible additional analyses according to the needs of main detailed project. Due to the size of the facility the need of filters for particulates is expected in systems V-1 and V-2, as well as revision of the system V-4. Adsorption filters are not needed in case of operation under regular working conditions. It would be favorable, but not indispensable to construct an adsorption filter that could be used occasionally at the input of V-2 system under the upper water shielding. This filter would be switched on when dehermetization of the reactor is planned and increased radioactivity is indicated in the gas system or experimental space of the RA reactor as well as in case of fuel elements failure. The aim of applying filters is shortening the time these elements would spend in the hermetized reactor i.e. their fast removal from the reactor. It is indispensable to build-in 'emergency' highly efficient adsorption filters in the V-2 system which would be switched on in case of accidents that would cause fuel elements meltdown (with simultaneous accident dehermetization of the heavy water system) [sr

The ICRH tokamak fusion test reactor

International Nuclear Information System (INIS)

Perkins, F.W.

1976-01-01

A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

X-ray digital industrial radiography (DIR) for local liquid velocity (V{sub LL}) measurement in trickle bed reactors (TBRs): Validation of the technique

Energy Technology Data Exchange (ETDEWEB)

Mohd Salleh, Khairul Anuar, E-mail: kmfgf@mst.edu; Lee, Hyoung Koo [Department of Mining and Nuclear Engineering, Missouri University of Science and Technology, Fulton Hall, 310 W. 14th St., Rolla, Missouri 65409 (United States); Rahman, Mohd Fitri Abdul [Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 143 Schrenk Hall, 400 W. 11th St., Rolla, Missouri 65409 (United States); Al Dahhan, Muthanna H. [Department of Mining and Nuclear Engineering, Missouri University of Science and Technology, Fulton Hall, 310 W. 14th St., Rolla, Missouri 65409 (United States); Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 143 Schrenk Hall, 400 W. 11th St., Rolla, Missouri 65409 (United States)

2014-06-15

Local liquid velocity measurements in Trickle Bed Reactors (TBRs) are one of the essential components in its hydrodynamic studies. These measurements are used to effectively determine a reactor's operating condition. This study was conducted to validate a newly developed technique that combines Digital Industrial Radiography (DIR) with Particle Tracking Velocimetry (PTV) to measure the Local Liquid Velocity (V{sub LL}) inside TBRs. Three millimeter-sized Expanded Polystyrene (EPS) beads were used as packing material. Three validation procedures were designed to test the newly developed technique. All procedures and statistical approaches provided strong evidence that the technique can be used to measure the V{sub LL} within TBRs.

Refurbishing the BR2 materials testing reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Dekeyser, J.; Gubel, P.

1995-01-01

SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

Analysis of the ZPPR-15 Critical Experiments with ENDF/B-V.2 and ENDF/B-VII.0 Data

International Nuclear Information System (INIS)

Kim, Sang Ji; Yang, Won Sik; Lee, Changho

2008-01-01

This paper presents the analysis results for the ZPPR-15 critical experiments. Using the ENDF/B-V.2 and ENDF/B-VII.0 data, three loading configurations of the ZPPR-15 Phase A experiments were analyzed with the ANL code suite for a fast reactor neutronics analysis, including the recently updated MC 2 -2 code. For the VIM Monte Carlo analyses with 3-D as-built models, the ENDF/B-VII.0 data improved the core multiplication factors by 0.21 to 0.37 %Δk, relative to the ENDF/B-V.2 data. With the plate heterogeneity effects taken into account by the SDX 1-D unit cell calculations, the DIF3D nodal transport solutions with the ENDF/B-V.2 data showed a good agreement for the core multiplication factors with the VIM Monte Carlo results to within 0.12 %Δk, but those with the ENDF/B-VII.0 data showed relatively larger deviations. Sensitivity studies based on the RZ models with homogenized cells showed excellent agreement for the core multiplication factors between the deterministic and Monte Carlo calculations to within 0.1 %Δk for both ENDF/B data. These results indicate that the MC 2 -2 methods are adequate for generating the multigroup cross sections for a fast reactor analysis, but the SDX process to account for the heterogeneity effect needs to be improved for the ENDF/B-VII.0 data. (authors)

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

International Nuclear Information System (INIS)

1966-01-01

This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

SIRIUS 2: A versatile medium power research reactor

International Nuclear Information System (INIS)

Rousselle, P.

1992-01-01

Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

On the anisotropy of in-reactor creep of Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Causey, A.R.; Holt, R.A.

1993-06-01

Creep specimens made from cold-worked Zr-2.5Nb tubes, fabricated with two different microstructures and crystallographic textures, were irradiated in the Osiris reactor in France in a fast-neutron flux of about 1.8 x 10 18 n.m -2 .s -1 , E > MeV, at 553 and 585 K. The hoop stresses from internal Fluences, up to 4 x 10 25 n.m -2 , more than double those achieved an any other creep test on cold-worked Zr-2.5Nb in which both axial and transverse strain were measured. Creep rates were obtained from strain versus fluence plots, and creep compliances were obtained from plots of the strain rates against hoop stress for each material at each temperature. The ratio of creep rates at 583 K to those at 553 K was ∼ 1.36, a little higher than that extrapolated from stress relaxation results at temperatures between 523 and 568 K. The ratio of the biaxial creep compliance in the axial direction to that in the transverse directions is different for the two test materials: 0.0 to -0.1 for the fuel sheathing texture and 0.5 to 0.6 for the pressure tube texture. The results were analysed using a self-consistent model developed to account for the contributions to the creep anisotropy of the three microstructure parameters involved and to account for the grain interaction effects. The model, which was normalized to test reactor and power reactor creep data for cold-worked Zr-2.5Nb tubes, predicted the ratio of the creep compliancies to be -0.26 and 0.63, respectively. Thus the creep anisotropy of Zr-2.5Nb tubes with pressure-tube-like crystallographic texture can be adequately predicted. (author). 18 refs., 4 tabs., 13 figs

Improvement of operational performance and increase of safety of WWER-1000/V-392

International Nuclear Information System (INIS)

Kurakov, Y.A.; Dragunov, Y.G.; Podshibiakin, A.K.; Fil, N.S.; Krushelnitsky, V.N.; Berkovich, V.M.

2001-01-01

The national programme of nuclear power development approved by the Russian Federation Government in 1998 considers the design of WWER-1000/V-392 power unit as a priority project of the new generation NPP with improved operational performances and increased safety. The pilot unit of this design (NVAES-2) is licensed for construction at the Novovoronezh NPP site. The NVAES-2 design is developed on the basis of standard power unit with reactor plant V-320. Twenty units of this type are in operation at the nuclear power plants in Russia, Ukraine and Bulgaria having totally about 270 reactor-years of operation. Two more V-320 units are being commissioned this year at Rostov NPP and Temelin NPP. So, the WWER-1000/V-392 design is as a whole an evolutionary development of the operating standard unit WWER-1000/V-320. Many technical solutions aimed at increase of safety and improvement of operational performance of the plant are implemented in the NVAES-2 design, such as advanced reactor WWER-1000, passive system of residual power removal, passive system of the core flooding under loss-of-coolant accidents, and others. NVAES-2 design refers to a class of advanced light water reactors and corresponds to the international requirements imposed to the nuclear power plants to be put into operation after the year 2000. New V-392 power unit has a good perspective from the view point of extensive implementation in the framework of the nuclear electricity production in Russia. Design decisions on NVAES-2 power unit with WWER-1000/V-392 reactor plant which assure significantly higher safety level and improve economical performance as compared to the operating WWER-1000 units are briefly considered in the present paper. (author)

Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, V Deo, Odredjivanje fine raspodele fluksa

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-07-15

Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S{sub n} code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S{sub 4} approximation. Reactor cell was treated as D{sub 2}O-U-D{sub 2}O system. Time of iteration was 185 s. Resena je transportna monoenergetska jednacina numerickom metodom Carlsona u cilindricnoj geometriji. Razvijanje S{sub n} kod za digitalnu masinu ZUSE-Z 23. Odredjena je raspodela fluksa u celiji reaktora RA S{sub 4} aproksimacijom. Celija je tretirana kao D{sub 2}O-U-D{sub 2}O. Vreme iteracije je 185 sec (author)

OTUS - Reactor inventory management system based on ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

1995-10-01

ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

The 2nd reactor core of the NS Otto Hahn

International Nuclear Information System (INIS)

Manthey, H.J.; Kracht, H.

1979-01-01

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

The influence of Triga 2000 reactor operation on the surface contamination at reactor room using smear test method

International Nuclear Information System (INIS)

Bintu Khoiriyyah; Budi Purnama; Tri Cahyo Laksono

2016-01-01

The monitoring of surface contamination should be conducted to determine the safety of work areas. Surface contamination at the TRIGA 2000 reactor room which is on PSTNT-BATAN Bandung remain to be implemented although reactor not operating. In this research monitoring of surface contamination when TRIGA 2000 in operation of the first time after several years not operating aims to determine the influence on the results of monitoring. The monitoring of surface contamination has been done using smear test method at some predetermined in TRIGA 2000 reactor room. The highest surface contamination activities is obtained 0.32 Bq/cm 2 and there are some points that are not detected. Based on keputusan kepala BAPETEN No.1/Ka BAPETEN/ V/99 the work showed that the TRIGA 2000 reactor in the category of low area contamination, that is <3.7 Bq/cm 2 to gross beta. (author)

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1977-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80%. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59% and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high re-circulating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)]. By contrast, the fusion-fission reactor design is not penalized by re-circulating power and uses relatively near-term fusion technology being developed for the fusion power program. New results are presented on the Th- 233 U and the U- 239 Pu fuel cycles. The purpose of this hybrid is fuel production, with projected costs at $55/g of Pu or $127/g of 233 U. Blanket and cooling system designs, including an emergency cooling system, by General Atomic Company, lead us to the opinion that the reactor can meet expected safety standards for licensing. The smallest mirror reactor having only a shield between the plasma and the coil is the 4.2-m long fusion engineering research facility (FERF) designed for material irradiation. The smallest mirror reactor having both a blanket and shield is the 7.5-m long experimental power reactor (EPR), which has both a fusion and a fusion-fission version. (author)

Fission product release from UO2 during irradiation. Diffusion data and their application to reactor fuel pins

International Nuclear Information System (INIS)

Findlay, J.R.; Johnson, F.A.; Turnbull, J.A.; Friskney, C.A.

1980-01-01

Release of fission product species from UO 2 , and to a limited extent from (U, Pu)0 2 was studied using small scale in-reactor experiments in which these interacting variables may be separated, as far as is possible, and their influences assessed. Experiments were at fuel ratings appropriate to water reactor fuel elements and both single crystal and poly-crystalline specimens were used. They employed highly enriched uranium such that the relative number of fissions occurring in plutonium formed by neutron capture was small. The surface to volume ratio (S/V) of the specimens was well defined thus reducing the uncertainties in the derivation of diffusion coefficients. These experiments demonstrate many of the important characteristics of fission product behaviour in UO 2 during irradiation. The samples used for these experiments were small being always less than 1g with a fissile content usually between 2 and 5mg. Polycrystalline materials were taken from batches of production fuel prepared by conventional pressing and sintering techniques. The enriched single crystals were grown from a melt of sodium and potassium chloride doped with UO 2 powder 20% 235 U content. The irradiations were performed in the DIDO reactor at Harwell. The neutron flux at the specimen was 4x10 16 neutrons m -2 s -1 providing a heat rating within the samples of 34.5 MW/teU

Design project of the dosimetry control system in the independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. V

International Nuclear Information System (INIS)

1964-01-01

Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels includes the following: calculations of CO 2 gas activity, design of the dosimetry control system, review of the changes that should be done in the RA reactor building for installing the independent CO 2 loop, specification of the materials with cost estimation, engineering drawings of the system [sr

Sequential injection spectrophotometric determination of V(V) in ...

African Journals Online (AJOL)

2009-01-15

Jan 15, 2009 ... The determination was based on the oxidation of dopamine (DP) by V(V) in acidic medium followed by ... To overcome some of these drawbacks some flow injec- ..... The analytical signal increased with the reactor length.

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

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

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

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

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Innovative energy production in fusion reactors

Science.gov (United States)

Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

Calculation of induced activity in the V-230 reactor

International Nuclear Information System (INIS)

Bouhahhane, A.; Farkas, G.

2013-01-01

In this paper, we focused on the calculation of the neutron induced activity of nuclear reactor components for decommissioning purposes. The results confirm, that the most important radionuclides in the reactor components dismantling process are 55 Fe (1 st decade), 60 Co (10 - 50 y) and 63 Ni (during the whole process). Another aim of this paper was to refer to the possibility to improve the accuracy of the calculations using continuous energy Monte Carlo methods. (authors)

The Jules Horowitz reactor project, a driver for revival of the research reactor community

Energy Technology Data Exchange (ETDEWEB)

Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

2010-07-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

Technology requirements for fusion--fission reactors based on magnetic-mirror confinement

International Nuclear Information System (INIS)

Moir, R.W.

1978-01-01

Technology requirements for mirror hybrid reactors are discussed. The required 120-keV neutral beams can use positive ions. The magnetic fields are 8 T or under and can use NbTi superconductors. The value of Q (where Q is the ratio of fusion power to injection power) should be in the range of 1 to 2 for economic reasons relating to the cost of recirculating power. The wall loading of 14-MeV neutrons should be in the range of 1 to 2 MW/m 2 for economic reasons. Five-times higher wall loading will likely be needed if fusion reactors are to be economical. The magnetic mirror experiments 2XIIB, TMX, and MFTF are described

Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

International Nuclear Information System (INIS)

Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

2010-01-01

The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

Very high flux steady state reactor and accelerator based sources

International Nuclear Information System (INIS)

Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

2004-01-01

With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

Power cables with extruded insulation and their accessories for rated voltages from 1 kV (Um = 1,2 kV) up to 30 kV (Um = 36 kV) : Part 2: cables for rated voltages from 6 kV (Um = 7,2 kV) up to 30 kV (Um = 36 kV)

CERN Document Server

International Electrotechnical Commission. Geneva

2005-01-01

Power cables with extruded insulation and their accessories for rated voltages from 1 kV (Um = 1,2 kV) up to 30 kV (Um = 36 kV) : Part 2: cables for rated voltages from 6 kV (Um = 7,2 kV) up to 30 kV (Um = 36 kV)

Tokamak engineering test reactor

International Nuclear Information System (INIS)

Conn, R.W.; Jassby, D.L.

1975-07-01

The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

International Nuclear Information System (INIS)

2005-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Neuronal responses in visual area V2 (V2) of macaque monkeys with strabismic amblyopia.

Science.gov (United States)

Bi, H; Zhang, B; Tao, X; Harwerth, R S; Smith, E L; Chino, Y M

2011-09-01

Amblyopia, a developmental disorder of spatial vision, is thought to result from a cascade of cortical deficits over several processing stages beginning at the primary visual cortex (V1). However, beyond V1, little is known about how cortical development limits the visual performance of amblyopic primates. We quantitatively analyzed the monocular and binocular responses of V1 and V2 neurons in a group of strabismic monkeys exhibiting varying depths of amblyopia. Unlike in V1, the relative effectiveness of the affected eye to drive V2 neurons was drastically reduced in the amblyopic monkeys. The spatial resolution and the orientation bias of V2, but not V1, neurons were subnormal for the affected eyes. Binocular suppression was robust in both cortical areas, and the magnitude of suppression in individual monkeys was correlated with the depth of their amblyopia. These results suggest that the reduced functional connections beyond V1 and the subnormal spatial filter properties of V2 neurons might have substantially limited the sensitivity of the amblyopic eyes and that interocular suppression was likely to have played a key role in the observed alterations of V2 responses and the emergence of amblyopia.

Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

International Nuclear Information System (INIS)

Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

2012-01-01

10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

International Nuclear Information System (INIS)

Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

2014-01-01

An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

Oxidation of SO{sub 2} in a trickle bed reactor packed with activated carbon at low liquid flow rates

Energy Technology Data Exchange (ETDEWEB)

Suyadal, Y.; Oguz, H. [Ankara Univ. (Turkey). Dept. of Chemical Engineering

2000-07-01

In this study, the oxidation of SO{sub 2} on activated carbon (AcC) by using distilled water and air was carried out in a laboratory scale trickle bed reactor (TBR). Distilled water and air containing 1.7% (v/v) SO{sub 2} were fed co-currently downward through a fixed bed of AcC particles in a range of 1-7 cm{sup 3}/s and 10-27 cm{sup 3}/s, respectively. H{sub 2}SO{sub 3}/H{sub 2}SO{sub 4} solutions were the products obtained in the liqiuid phase. Steady-state experiments were performed in a column of 0.15 m packing height and 0.047 m column diameter at 20 C and atmospheric pressure. (orig.)

Comparison of results for burning with BWR reactors CASMO and SCALE 6.2 (TRITON / NEWT); Comparacion de los resultados de quemado para reactores BWR con CASMO y SCALE 6.2 (TRITON/NEWT)

Energy Technology Data Exchange (ETDEWEB)

Mesado, C.; Miro, R.; Barrachina, T.; Verdu, G.

2014-07-01

In this paper we compare the results from two codes burned, CASMO and SCALE 6.2 (TRITON). To do this, is simulated all segments corresponding to a boiling water reactor (BWR) using both codes. In addition, to account for different working points, simulations changing the instantaneous variables, these are repeated: void fractions (6 points), fuel temperature (6 points) and control rods (two points), with a total of 72 possible combinations of different instantaneous variables for each segment. After all simulations are completed for each segment, we can reorder the obtained cross sections, as SCALE CASMO both, to create a library of compositions nemtab format. This format is accepted by the neutronic code of nodal diffusion, PARCS v2.7. Finally compares the results obtained with PARCS and with the SIMULATE3 -SIMTAB methodology to level of full reactor. Also, we have made use of the KENO-VI and MCDANCOFF modules belonging to SCALE. The first is a Monte Carlo transport code with which you can validate the value of the multiplier, the second has been used to obtain values of Dancoff factor and increase the accuracy of model SCALE. (Author)

NH{sub 3} decomposition and simultaneous H{sub 2} separation with a commercial Pd-Cu-Ag/V membrane

Energy Technology Data Exchange (ETDEWEB)

Skodras, G.; Sakelleropoulos, G. [Centre for Research and Technology, Hellas, Ptolemaidas-Kozanis, Ptolemaida (Greece). Inst. for Solid Fuel Technolgy and Applications]|[Aristotle Univ. of Thessaloniki, Thessaloniki (Greece). Dept. of Chemical Engineering, Chemical Process Engineering Lab]|[Centre for Research and Technology, Hellas, Thermis, Thessaloniki (Greece). Chemical Process Research Inst.; Kaldis, S. [Centre for Research and Technology, Hellas, Thermis, Thessaloniki (Greece). Chemical Process Research Inst.; Topis, S. [Centre for Research and Technology, Hellas, Ptolemaidas-Kozanis, Ptolemaida (Greece). Inst. for Solid Fuel Technolgy and Applications]|[Aristotle Univ. of Thessaloniki, Thessaloniki (Greece). Dept. of Chemical Engineering, Chemical Process Engineering Lab; Koutsonikolas, D. [Aristotle Univ. of Thessaloniki, Thessaloniki (Greece). Dept. of Chemical Engineering, Chemical Process Engineering Lab; Grammelis, P. [Centre for Research and Technology, Hellas, Ptolemaidas-Kozanis, Ptolemaida (Greece). Inst. for Solid Fuel Technolgy and Applications

2006-07-01

The potential for integrated gasification combined cycle (IGCC) technology to emerge as an efficient and environmentally clean technology for power generation from coal gas was discussed. Ammonia (NH{sub 3}) is formed during gasification of coal. The concentration in coal gas depends on temperature, pressure, residence time and coal rank. In the gas turbine as much as 50 per cent of the NH{sub 3} in the fuel gas can be converted to NOx when the gas is combusted to produce power. A catalyst is required to bring NH{sub 3} levels down to acceptable levels for a gas turbine. This study examined the simultaneous ammonia (NH{sub 3}) decomposition and hydrogen (H{sub 2}) separation via a commercial Pd-Cu-Ag/V catalytic membrane reactor with 100 per cent H{sub 2} selectivity. A 16 per cent Ni/Al{sub 2}O{sub 3} catalyst was prepared and 88 per cent NH{sub 3} conversion was achieved with 20 per cent H{sub 2} in the feed stream. Increase of temperature and prolongation of residence time resulted in higher conversions. However, pressure increase lowered the decomposition due to the high H{sub 2} partial pressure. In order to develop kinetic equations, experiments at differential conversions were also performed. H{sub 2} diffusion was found to be the rate limiting step of H{sub 2} transport through the membrane. As such, H{sub 2} permeance increased exponentially with temperature. Mathematical model were then developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. It was concluded that the conversion of NH{sub 3} can be increased significantly using the Pd-Cu-Ag/V membrane. 14 refs., 3 tabs., 7 figs.

Potential performance analysis and future trend prediction of electric vehicle with V2G/V2H/V2B capability

Directory of Open Access Journals (Sweden)

Dalong Guo

2016-03-01

Full Text Available Due to the intermittent nature, renewable energy sources (RES has brought new challenges on load balancing and energy dispatching to the Smart Grid. Potentially served as distributed energy storage, Electric Vehicle’s (EV battery can be used as a way to help mitigate the pressure of fluctuation brought by RES and reinforce the stability of power systems. This paper gives a comprehensive review of the current situation of EV technology and mainly emphasizing three EV discharging operations which are Vehicle to Grid (V2G, Vehicle to Home (V2H, and Vehicle to Building (V2B, respectively. When needed, EV’s battery can discharge and send its surplus energy back to power grid, residential homes, or buildings. Based on our data analysis, we argue that V2G with the largest transmission power losses is potentially less efficient compared with the other two modes. We show that the residential users have the incentive to schedule the charging, V2G, and V2H according to the real-time price (RTP and the market sell-back price. In addition, we discuss some challenges and potential risks resulting from EVs’ fast growth. Finally we propose some suggestions on future power systems and also argue that some incentives or rewards need to be provided to motivate EV owners to behave in the best interests of the overall power systems.

Operation and maintenance of nuclear power plant

International Nuclear Information System (INIS)

1990-04-01

The proceedings contain 34 papers, almost all discussing specific problems or the Dukovany nuclear power plant. The major part is devoted to the matters of maintenance, regular and unplanned repairs, decontamination of the steam generator for inspection and repair purposes, various methods of in-service diagnostics. Some papers discuss reactor start-up, safety assurance, unit control and economic aspects. Environmental impacts of the power plant are also tackled. Some special contributions concern corrosion, chemical analysis of the coolant and the diagnostics of electrical equipment. The possibility is discussed of switching WWER reactors to improved fuel cycles with increased fuel burnup. (M.D.). 37 figs., 23 tabs., 47 refs

Circuit designs for measuring reactor period, peak power, and pulse fluence on TRIGA and other pulse reactor

International Nuclear Information System (INIS)

Meyer, R.D.; Thome, F.V.; Williams, R.L.

1976-01-01

Inexpensive circuits for use in evaluating reactor pulse prompt period, peak power, and pulse fluence (NVT) are presented. In addition to low cost, these circuits are easily assembled and calibrated and operate with a high degree of accuracy. The positive period measuring system has been used in evaluating reactivity additions as small as 5 cents (with an accuracy of ±0.1 cents) and as large as $4.50 (accuracy ±2 cents). Reactor peak power is measured digitally with a system accuracy of ±0.04% of a 10 Volt input (±4 mV). The NVT circuit measures over a 2-1/2 decade range, has 3 place resolution and an accuracy of better than 1%. (author)

Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Villeger, S; Cousty, S; Ricard, A; Sixou, M

2003-01-01

Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

Neutron Energy Spectra from Neutron Induced Fission of 235U at 0.95 MeV and of 238U at 1.35 and 2.02 MeV

Energy Technology Data Exchange (ETDEWEB)

Almen, E; Holmqvist, B; Wiedling, T

1971-09-15

The shapes of fission neutron spectra are of interest for power reactor calculations. Recently it has been suggested that the neutron induced fission spectrum of 235U may be harder than was earlier assumed. For this reason measurements of the neutron spectra of some fissile isotopes are in progress at our laboratory. This report will present results from studies of the energy spectra of the neutrons emitted in the neutron induced fission of 235U and 238U. The measurements were performed at an incident neutron energy of 0.95 MeV for 235U and at energies of 1.35 and 2.02 MeV for 238U using time-of-flight techniques. The time-of-flight spectra were only analysed at energies higher than those of the incident neutrons and up to about 10 MeV. Corrections for neutron attenuation in the uranium samples were calculated using a Monte Carlo program. The corrected fission neutron spectra were fitted to Maxwellian temperature distributions. For 235U a temperature of 1.27 +- 0.01 MeV gives the best fit to the experimental data and for 238U the corresponding values are 1.29 +- 0.03 MeV at 1.35 MeV and 1.29 +- 0.02 MeV at 2.02 MeV

Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

Conversion of reactor neutrons by lithium deuteride; Konverzija reaktorskih neutrona pomocu litijumdeuterida

Energy Technology Data Exchange (ETDEWEB)

Strugar, P; Altiparmakov, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1979-07-01

Nuclear reactors are powerful neutron sources. But for many purposes, neutrons of higher than fission energy are needed. Such neutrons may be produced by the conversion of reactor neutrons into 14 MeV neutrons using lithium deuteride converter. For that converter is generally employed {sup 6}LiD, the usual material for the thermonuclear weapons, and therefore hardly accessible in needed quantity. That was the reason to analyse such converter made of LiD with the lithium of natural isotopic content. This analysis starts with the basic conversion relations, takes into account neutron absorption and tritium generation, and finally, estimates the 14 MeV neutron flux and the heat generated in proposed converter, when the converter was coupled to each of three Yugoslav nuclear reactors. Results show that the converter made of LiD with the natural lithium is 50% less efficient than the converter of {sup 6}LiD. Intensity of 14 MeV neutrons is within limits 5. 10{sup 5} - 10{sup 10} (n/cm{sup 2}.s) for the converter used either as external converter with reactor RB, within the thermal column in the reactor TRIGA or as a 'fuel' segment at the reactor RA. (author)

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

[Adsorption and removal of gas-phase Hg(0) over a V2O5/AC catalyst in the presence of SO2].

Science.gov (United States)

Wang, Jun-wei; Yang, Jian-li; Liu, Zhen-yu

2009-12-01

The adsorption and removal behaviors of gas-phase Hg(0) over V2O5/AC and AC were studied under a simulated flue gas (containing N2, SO2, O2) in a fixed-bed reactor. The influences of the V2O5, loading, SO2 concentration and adsorption temperature on Hg0 adsorption were investigated. The speciation of mercury adsorbed was determined by X-ray photoelectron spectroscopy (XPS). It was found that the V2O5/AC catalyst has a much higher capability than AC for Hg(0) adsorption and removal, mainly because of the catalytic oxidation activity of V2O5. The Hg(0) adsorption capability depends on the V2O5 content of the V2O5/AC catalyst. The amounts of mercury adsorbed increase from 75.9 microg x g(-1) to 89.6 microg x g(-1) (in the absence of O2) and from 115.9 microg x g(-1) to 185.5 microg x g(-1) (in the presence of O2) as the V2O5 loading increases from 0.5% to 1.0%, which are much higher than those over AC under the same conditions (9.6 microg x g(-1) and 23.3 microg x g(-1)). SO2 in the flue gas enhances Hg(0) adsorption over the V2O5/AC catalyst, which is due to the reaction of SO2 and Hg(0) on V2O3/AC. But as the SO2 concentration increases from 500 x 10(-6) to 2000 x 10(-6), the amount of mercury adsorbed has only a slight increase. The optimal temperature for Hg(0) adsorption over the V2O5/AC catalyst is around 150 degrees C, at which the amounts of mercury adsorbed are up to 98.5 microg x g(-1) (in the absence of O2) and 187.7 microg x g(-1) (in the presence of O2). The XPS results indicate the formation of Hg(0) and HgSO4 on the surface of the V2O5/AC catalyst, which confirms the role of V2O5 and SO2.

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results.

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

International Nuclear Information System (INIS)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

International Nuclear Information System (INIS)

Batista, J.L.; Renke, C.A.C.

1991-11-01

This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

Measurement and Calculation of Gamma Radiation from HWZPR Reactor

International Nuclear Information System (INIS)

Jalali, Majid

2006-01-01

HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)

Status and prospects of the core surveillance system SCORPIO-VVER in Czech Republic and Slovakia

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2011-01-01

The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (Czech Republic) and two units of Bohunice NPP (Slovak Republic) replacing the original Russian VK3 system. By both Czech and Slovak nuclear regulatory bodies it was licensed as a Technical Specification Surveillance tool. Since it's first installation, the development of SCORPIO-VVER system continues along with the changes in WWER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The latest most significant changes were done in connection with implementation of a new digital I and C system, loading of the optimized gadolinium bearing Gd2 fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions, in design and methodology of the limit and technical specifications checking (implementation of the on-line shutdown margin calculation to the system) and improvements in the predictive part of the system (Strategy Generator). After the currently finished upgrades (Upgrade 2 at EBO Slovakia in 2009, Upgrade 5 at EDU Czech Republic in 2010) the SCORPIO-VVER is still in focus of Central European nuclear power plants with the road map of modification and implementation up to 2015. In April 2011 the Upgrade 3 at EBO Slovakia has been contracted to support the changed reactor technical specification and for support of the new type of fuel planned to load in 2012. During the summer of 2011 the discussions started with EDU NPP in Czech Republic regarding to the future development of the SCORPIO-VVER system up to 2015. Parallel with the support of current installations at NPPs the project of new installations is ongoing. During

Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.2

International Nuclear Information System (INIS)

2014-01-01

The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Radionuclide distribution in TMI-2 reactor building basement liquids and solids

International Nuclear Information System (INIS)

Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

1984-01-01

As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

An experimental investigation of fission product release in SLOWPOKE-2 reactors

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

A comparison of the electrochemical recovery of palladium using a parallel flat plate flow-by reactor and a rotating cylinder electrode reactor

International Nuclear Information System (INIS)

Terrazas-Rodriguez, J.E.; Gutierrez-Granados, S.; Alatorre-Ordaz, M.A.; Ponce de Leon, C.; Walsh, F.C.

2011-01-01

The production of catalytic converters generates large amounts of waste water containing Pd 2+ , Rh 3+ and Nd 3+ ions. The electrochemical treatment of these solutions offers an economic and effective alternative to recover the precious metals in comparison with other traditional metal recovery technologies. The separation of palladium from this mixture of metal ions by catalytic deposition was carried out using a rotating cylinder electrode reactor (RCER) and a parallel plate reactor (FM01-LC) with the same cathode area (64 cm 2 ) and electrolyte volume (300 cm 3 ). The study was carried out at mean linear flow velocities of 1.27 -1 (120 e /v -1 (7390 2+ ions in the parallel plate electrode reactor was 35% while the recovery of 97% of Pd 2+ in the RCER was 62%. The volumetric energy consumption during the electrolysis was 0.56 kW h m -3 and 2.1 kW h m -3 for the RCER and the FM01-LC reactors, respectively. Using a three-dimensional stainless steel electrode in the FM01-LC laboratory reactor, 99% of palladium ions were recovered after 30 min of electrolysis while in the RCER, 120 min were necessary.

The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

International Nuclear Information System (INIS)

Hurt, S.S.

1986-11-01

The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

Evolution of ASTEC V1.2 rev.1 code for WWER-1000 reactors/SBO sequence

International Nuclear Information System (INIS)

Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Kalchev, B.; Passalacqua, R.

2006-01-01

In this paper a comparison between calculations of severe accidents occurred from WWER-1000 with ASTEC code specified for an event of full unloading with relief valves stuck opened with no hydroaccumulators intervention is presented. The purpose of the analyses provided is to present the relationship between the improvements of the actual version (ASTEC Vl.2 rev. 1) and ASTEC V1.1 p2 like: code modifications, incoming data improvements. Such discrepancies are to be examined. Case by case suggestions for ASTEC improvements are to be provided

Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

International Nuclear Information System (INIS)

Chatt, A.; Holzbecher, J.

1990-01-01

SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

International Nuclear Information System (INIS)

Atanasova, B.; Grudev, P.

2011-01-01

The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

Preliminary Options Assessment of Versatile Irradiation Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

International Nuclear Information System (INIS)

Richmond, R.

1982-12-01

The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

International Nuclear Information System (INIS)

Greene, S.R.; Spellman, D.J.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

Crowdsourcing-Assisted Radio Environment Database for V2V Communication

Directory of Open Access Journals (Sweden)

Keita Katagiri

2018-04-01

Full Text Available In order to realize reliable Vehicle-to-Vehicle (V2V communication systems for autonomous driving, the recognition of radio propagation becomes an important technology. However, in the current wireless distributed network systems, it is difficult to accurately estimate the radio propagation characteristics because of the locality of the radio propagation caused by surrounding buildings and geographical features. In this paper, we propose a measurement-based radio environment database for improving the accuracy of the radio environment estimation in the V2V communication systems. The database first gathers measurement datasets of the received signal strength indicator (RSSI related to the transmission/reception locations from V2V systems. By using the datasets, the average received power maps linked with transmitter and receiver locations are generated. We have performed measurement campaigns of V2V communications in the real environment to observe RSSI for the database construction. Our results show that the proposed method has higher accuracy of the radio propagation estimation than the conventional path loss model-based estimation.

Dynamic simulation of the 2 MWt slowpoke heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1982-04-01

A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

Distribution of energy of impulses of the modernized IBR-2 REACTOR

International Nuclear Information System (INIS)

Tayibov, L.A; Mehtiyeva, R.N.; )

2011-01-01

Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

Commissioning of the Opal reactor cold neutron source

International Nuclear Information System (INIS)

Thiering, R.; Lu, W.; Ullah, R.

2006-01-01

Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies ( 1 0 m eV). A cold neutron flux of 1.4 1 0 E 1 4 n /cm 2/ s is expected, with a peak in the energy spectrum at 4.2m eV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24 K. The moderator chamber, which contains the deuterium, has been constructed from AlMg 5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 3 00 K ), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

Rapid restoration of methanogenesis in an acidified UASB reactor treating 2,4,6-trichlorophenol (TCP).

Science.gov (United States)

Díaz-Báez, María Consuelo; Valderrama-Rincon, Juan Daniel

2017-02-15

Anaerobic bioreactors are often used for removal of xenobiotic and highly toxic pollutants from wastewater. Most of the time, the pollutant is so toxic that the stability of the reactor becomes compromised. It is well known that methanogens are one of the most sensitive organisms in the anaerobic consortia and hence the stability of the reactors is highly dependant on methanogenesis. Unfortunately few studies have focused on recovering the methanogenic activity once it has been inhibited by highly toxic pollutants. Here we establish a quick recovery strategy for neutralization of an acidified UASB reactor after failure by intoxication with an excess of TCP in the influent. Once the reactor returned to pH values compatible with methanogenesis, biogas production was re-started after one day and the system was re-acclimated to TCP. Successful removal of TCP from synthetic wastewater was shown for concentrations up to 70mg/L after restoration. Copyright © 2016 Elsevier B.V. All rights reserved.

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

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

A conceptual design of LIB fusion reactor: UTLIF(2)

International Nuclear Information System (INIS)

Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

1984-01-01

UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

Directory of Open Access Journals (Sweden)

Yasmine Abdel-Maksoud

2016-09-01

Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

Beam Tracking in Switched-Beam Antenna System for V2V Communication

Directory of Open Access Journals (Sweden)

Settawit Poochaya

2016-01-01

Full Text Available This paper presents the proposed switched beam antenna system for V2V communication including optimum antenna half power beamwidth determination in urban road environments. SQP optimization method is selected for the computation of optimum antenna half power beamwidth. In addition, beam tracking algorithm is applied to guarantee the best beam selection with maximum RSSI. The results present the success of the proposed system with the increasing of V2V performance metrics. Also, V2V data dissemination via the proposed system introduces the enhancement of V2V link in terms of RSSI, PER, BER, Tsafe, and Rsafe. The results indicate the improvement of V2V link reliability. Consequently, the road safety is improved.

Determination of buckling in the IPEN/MB-01 Reactor in cylindrical configuration

Energy Technology Data Exchange (ETDEWEB)

Purgato, Rafael Turrini; Bitelli, Ulysses d' Utra; Aredes, Vitor Ottoni; Silva, Alexandre F. Povoa da; Santos, Diogo Feliciano dos; Mura, Luis Felipe L.; Jerez, Rogerio, E-mail: rtpurgato@ipen.br, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2013-07-01

One of the key parameters in reactor physics is the buckling of a reactor core. It is related to important parameters such as reaction rates, nuclear power operation, fuel burning, among others. In a critical reactor, the buckling depends on the geometric and material characteristics of the reactor core. This paper presents the results of experimental buckling in the reactor IPEN/MB-01 nuclear reactor in its cylindrical configuration with 28 fuel rods along its diameter. The IPEN/MB-01 is a zero power reactor designed to operate at a maximum power of 100 watts, it is a versatile nuclear facility which allows the simulation of all the characteristics of a large nuclear power reactor and ideal for this type of measurement. We conducted a mapping of neutron flux inside the reactor and thereby determined the total buckling of the cylindrical configuration. The reactor was operated for an hour. Then, the activation of the fuel rods was measured by gamma spectrometry on a rod scanner HPGe detector. We analyzed the gamma photons of the {sup 239}Np (276,6 keV) for neutron capture and the {sup 143}Ce (293,3 keV) for fission on both {sup 238}U and {sup 235}U, respectively. We analyzed the axial and radial directions. Other measurements were performed using wires and gold foils in the radial and axial directions of the reactor core. The results showed that the cylindrical configuration compared to standard rectangular configuration of the IPEN/MB-01 reactor has a higher neutron economy, since in this configuration there is less leakage of neutrons. The Buckling Total obtained from the three methods was 95.84 ± 2.67 m{sup -2}. (author)

Methodology for the evaluation of tolerability of defects in WWER-1000/V-320 reactor pressure vessels

International Nuclear Information System (INIS)

Brumovsky, M.; Horacek, L.; Ruscak, M.

1996-05-01

The methodology provides guidelines for the assessment of tolerability of defects found during in-service inspection of the base material and overlay of WWER-1000/V-320 type reactor pressure vessels. With regard to the method of calculating the tolerability of defects and rules for the preparation and implementation of repairs, this methodology can also find use in the assessment of tolerability of defects in selected facilities of WWER-1000/V-320 type nuclear power plants provided that adequate input data concerning the materials, manufacturing technology, and operating load regime are available and that the facilities are made of ferrite/bainite type steels. This methodology should serve as a binding document underlying the development of a technical approach to provisions for a further operation of facilities in which intolerable defects have been found by nondestructive testing. (author)

Congestion based mechanism for route discovery in a V2I-V2V system applying smart devices and IoT.

Science.gov (United States)

Parrado, Natalia; Donoso, Yezid

2015-03-31

The Internet of Things is a new paradigm in which objects in a specific context can be integrated into traditional communication networks to actively participate in solving a determined problem. The Vehicle-to-Vehicle (V2V) and Vehicle-to-Infrastructure (V2I) technologies are specific cases of IoT and key enablers for Intelligent Transportation Systems (ITS). V2V and V2I have been widely used to solve different problems associated with transportation in cities, in which the most important is traffic congestion. A high percentage of congestion is usually presented by the inappropriate use of resources in vehicular infrastructure. In addition, the integration of traffic congestion in decision making for vehicular traffic is a challenge due to its high dynamic behavior. In this paper, an optimization model over the load balancing in the congestion percentage of the streets is formulated. Later, we explore a fully congestion-oriented route discovery mechanism and we make a proposal on the communication infrastructure that should support it based on V2I and V2V communication. The mechanism is also compared with a modified Dijkstra's approach that reacts at congestion states. Finally, we compare the results of the efficiency of the vehicle's trip with the efficiency in the use of the capacity of the vehicular network.

Congestion Based Mechanism for Route Discovery in a V2I-V2V System Applying Smart Devices and IoT

Directory of Open Access Journals (Sweden)

Natalia Parrado

2015-03-01

Full Text Available The Internet of Things is a new paradigm in which objects in a specific context can be integrated into traditional communication networks to actively participate in solving a determined problem. The Vehicle-to-Vehicle (V2V and Vehicle-to-Infrastructure (V2I technologies are specific cases of IoT and key enablers for Intelligent Transportation Systems (ITS. V2V and V2I have been widely used to solve different problems associated with transportation in cities, in which the most important is traffic congestion. A high percentage of congestion is usually presented by the inappropriate use of resources in vehicular infrastructure. In addition, the integration of traffic congestion in decision making for vehicular traffic is a challenge due to its high dynamic behavior. In this paper, an optimization model over the load balancing in the congestion percentage of the streets is formulated. Later, we explore a fully congestion-oriented route discovery mechanism and we make a proposal on the communication infrastructure that should support it based on V2I and V2V communication. The mechanism is also compared with a modified Dijkstra’s approach that reacts at congestion states. Finally, we compare the results of the efficiency of the vehicle’s trip with the efficiency in the use of the capacity of the vehicular network.

Planned Scientific programs around the Triga Mark 2 Reactor

International Nuclear Information System (INIS)

Majah, M Ibn.

2007-01-01

Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

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

International Nuclear Information System (INIS)

Gauld, I.C.

2000-01-01

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

Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

International Nuclear Information System (INIS)

Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak; Park, Young Soo

2010-01-01

The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

Energy Technology Data Exchange (ETDEWEB)

Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Park, Young Soo [Korea Hydro and Nuclear Power Co., Busan (Korea, Republic of)

2010-10-15

The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

Beam plasma 14 MeV neutron source for fusion materials development

International Nuclear Information System (INIS)

Ravenscroft, D.; Bulmer, D.; Coensgen, F.; Doggett, J.; Molvik, A.; Souza, P.; Summers, L.; Williamson, V.

1991-09-01

The conceptual engineering design and expected performance for a 14 MeV DT neutron source is detailed. The source would provide an intense neutron flux for accelerated testing of fusion reactor materials. The 150-keV neutral beams inject energetic deuterium atoms, that ionize, are trapped, then react with a warm (200 eV), dense tritium target plasma. This produces a neutron source strength of 3.6 x 10 17 n/sec for a neutron power density at the plasma edge of 5--10 MW/m 2 . This is several times the ∼2 MW/m 2 anticipated at the first wall of fusion reactors. This high flux provides accelerated end-of-life tests of 1- to 2-year duration, thus making materials development possible. The modular design of the source and the facilities are described

Power noise spectrum classification in the problem of the IBR-2 reactor

International Nuclear Information System (INIS)

Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

1988-01-01

The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

International Nuclear Information System (INIS)

Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

1987-01-01

Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

Research reactor FR2 - 20 years chemical and radiochemical measurements

International Nuclear Information System (INIS)

Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

1986-09-01

The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

Venting krypton-85 from the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Burton, H.M.

1981-01-01

To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

Research on economics and CO2 emission of magnetic and inertial fusion reactors

International Nuclear Information System (INIS)

Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

2011-01-01

An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

Physics of plutonium recycling: volume V. Plutonium recycling in fast reactors

International Nuclear Information System (INIS)

1996-01-01

As part of a programme proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this report, the multi-recycle performance of the metal-fuelled benchmark is evaluated. Benchmark results assess the reactor performance and toxicity behaviour in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilised to significantly reduce the radiotoxicity originating in the LWR cycle which would otherwise be destined for burial. (Author). tabs., figs., refs

Development of a TiO2-coated optical fiber reactor for water decontamination

International Nuclear Information System (INIS)

Danion, A.

2004-09-01

The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

A comparison of mass transfer coefficients between trickle-bed, hollow fiber membrane and stirred tank reactors.

Science.gov (United States)

Orgill, James J; Atiyeh, Hasan K; Devarapalli, Mamatha; Phillips, John R; Lewis, Randy S; Huhnke, Raymond L

2013-04-01

Trickle-bed reactor (TBR), hollow fiber membrane reactor (HFR) and stirred tank reactor (STR) can be used in fermentation of sparingly soluble gasses such as CO and H2 to produce biofuels and bio-based chemicals. Gas fermenting reactors must provide high mass transfer capabilities that match the kinetic requirements of the microorganisms used. The present study compared the volumetric mass transfer coefficient (K(tot)A/V(L)) of three reactor types; the TBR with 3 mm and 6 mm beads, five different modules of HFRs, and the STR. The analysis was performed using O2 as the gaseous mass transfer agent. The non-porous polydimethylsiloxane (PDMS) HFR provided the highest K(tot)A/V(L) (1062 h(-1)), followed by the TBR with 6mm beads (421 h(-1)), and then the STR (114 h(-1)). The mass transfer characteristics in each reactor were affected by agitation speed, and gas and liquid flow rates. Furthermore, issues regarding the comparison of mass transfer coefficients are discussed. Copyright © 2013 Elsevier Ltd. All rights reserved.

Computational investigation of 99Mo, 89Sr, and 131I production rates in a subcritical UO2(NO32 aqueous solution reactor driven by a 30-MeV proton accelerator

Directory of Open Access Journals (Sweden)

Z. Gholamzadeh

2015-12-01

Full Text Available The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing 99Mo. In this method, the medical isotope production system itself is used to extract 99Mo or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of 99Mo by irradiating targets. In this study, the neutronic performance and 99Mo, 89Sr, and 131I production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ∼1,500 Ci/wk (∼325 6-day Ci of 99Mo at the end of a cycle.

Application of ASTEC V2.0 to severe accident analyses for German KONVOI type reactors

International Nuclear Information System (INIS)

Nowack, H.; Erdmann, W.; Reinke, N.

2011-01-01

The integral code ASTEC is jointly developed by IRSN (Institut de Radioprotection et de Surete Nucleaire, France) and GRS (Germany). Its main objective is to simulate severe accident scenarios in PWRs from the initiating event up to the release of radioactive material into the environment. This paper describes the ASTEC modeling approach and the nodalisation of a KONVOI type PWR as an application example. Results from an integral severe accident study are presented and shortcomings as well as advantages are outlined. As a conclusion, the applicability of ASTEC V2.0 for deterministic severe accident analyses used for PSA level 2 and Severe Accident Management studies will be assessed. (author)

2-minute Gridded Global Relief Data (ETOPO2) v2

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — Two-minute gridded global relief for both ocean and land areas are available in the ETOPO2v2 (2006) database. ETOPO2v2 replaced ETOPO2 (2001). The historic 2-minute...

TMI-2 reactor-vessel head removal and damaged-core-removal planning

International Nuclear Information System (INIS)

Logan, J.A.; Hultman, C.W.; Lewis, T.J.

1982-01-01

A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results.

Czech safety concept: 2013 state of the art

International Nuclear Information System (INIS)

Vokal, Antonin; Slovak, Jiri

2014-01-01

The Czech Republic operates four WWER 440 reactors (Dukovany) and two WWER 1000 reactors (Temelin). The four 440 MW Dukovany units were installed and began operation during the period 1985-1988. The two WWER 1000 reactors at Temelin started operation in 2002 and 2003. Currently, more than 8000 SF assemblies from WWER 440 reactors and 900 spent assemblies from WWER 1000 reactors spent fuel assemblies are stored in dry storage facilities located in the area of both NPP in approved casks or in pools at reactor sites. More than 4 000 assemblies are expected to be spent by 2025 at Dukovany reactors and 4 600 assemblies by 2042 at Temelin reactors. The multi-billion Euro contract to build two new nuclear reactors at the current site of Temelin with the option for an additional one in Dukovany has recently been launched in the Czech Republic. It is expected that more than 8 000 fuel assemblies would be spent in the three new nuclear reactors in the Czech Republic during their 60 years of electricity production. The basic reference plan is to directly dispose of all of the spent fuel assemblies in a deep geological repository (DGR), starting operation not earlier than in 2065. The DGR is planned to be located in granite host rock, because no other type of host rock in sufficient volume is available in the Czech Republic. Currently seven candidate sites for DGR suitable for geological disposal of SF assemblies have been selected, but due to negative community attitudes at the notion of have a repository in their backyard, they are still awaiting a detailed geological survey. According to proposed reference designs, SF assemblies should be in steel-based canisters emplaced in vertical or horizontal boreholes in granite host rock at approximately 500 m under the surface and surrounded by compacted bentonite. The Czech safety concept is based on the KBS-3 concept developed in Sweden. The Swedish concept is primarily based on almost thermodynamic stability of copper overpack in

Neutron spectra in two beam ports of the TRIGA Mark III reactor

International Nuclear Information System (INIS)

Vega C, H. R.; Hernandez D, V. M.; Aguilar, F.; Paredes, L.; Rivera M, T.

2013-10-01

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6 Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

Neutron spectra in two beam ports of the TRIGA Mark III reactor

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

2013-10-15

The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

CO{sub 2} Energy Reactor – Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

Energy Technology Data Exchange (ETDEWEB)

Santos, Rafael M., E-mail: rafael.santos@alumni.utoronto.ca [Chemical and Environmental Laboratories (CEL), School of Applied Chemical and Environmental Sciences, Sheridan Institute of Technology, Brampton, ON (Canada); Knops, Pol C. M.; Rijnsburger, Keesjan L. [Innovation Concepts B.V., Twello (Netherlands); Chiang, Yi Wai [School of Engineering, University of Guelph, Guelph, ON (Canada)

2016-02-15

To overcome the challenges of mineral CO{sub 2} sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In this paper, a perspective on the status of this technology and outlook for the future is provided. To date, laboratory-scale tests of the envisioned process have been performed in a tubular “rocking autoclave” reactor. The mineral of choice has been olivine [~Mg{sub 1.6}Fe{sup 2+}{sub 0.4}(SiO{sub 4}) + ppm Ni/Cr], although asbestos, steel slags, and oil shale residues are also under investigation. The effect of several process parameters on reaction extent and product properties has been tested: CO{sub 2} pressure, temperature, residence time, additives (buffers, lixiviants, chelators, oxidizers), solids loading, and mixing rate. The products (carbonates, amorphous silica, and chromite) have been physically separated (based on size, density, and magnetic properties), characterized (for chemistry, mineralogy, and morphology), and tested in intended applications (as pozzolanic carbon-negative building material). Economically, it is found that product value is the main driver for mineral carbonation, rather than, or in addition to, the sequestered CO{sub 2}. The approach of using a GPV and focusing on valuable reaction products could thus make CO{sub 2} mineralization a feasible and sustainable industrial process.

Safety and upgrading of nuclear power plants with WWER-440/V-230; Voprosy bezopasnosti i modernizatsii AEhS s WWER-440 (V-230)

Energy Technology Data Exchange (ETDEWEB)

Melnikov, N [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

1996-12-31

Modernization of WWER-440/V-230 reactors is considered according to the WANO project `Proposal for reconstruction of WWER-440/V-230 reactors` (1990). The main technical measures to be taken relate to: 1. Expanding the considered spectrum of design basis accidents; 2. Provision of first loop integrity; 3. Safety system structure and components reliability enhancement; 4. Reduction o initiating events probability; 5. Enhancement of hermeticity and integrity of the area surrounding the first loop; 6. Fire fighting system upgrade; 7. Radiation control system upgrade; 8. Provision of seismic resistance. Details of the recommended technical steps for some NPPs are presented. 1 tab.

Measurement of cross sections for the scattering of neutrons in the energy range from 2 MeV to 4 MeV with the {sup 15}N(p,n) reaction as neutron source; Messung von Wirkungsquerschnitten fuer die Streuung von Neutronen im Energiebereich von 2 MeV bis 4 MeV mit der {sup 15}N(p,n)-Reaktion als Neutronenquelle

Energy Technology Data Exchange (ETDEWEB)

Poenitz, Erik

2010-04-26

In future nuclear facilities, the materials lead and bismuth can play a more important role than in today's nuclear reactors. Reliable cross section data are required for the design of those facilities. In particular the neutron transport in the lead spallation target of an Accelerator-Driven Subcritical Reactor strongly depends on the inelastic neutron scattering cross sections in the energy region from 0.5 MeV to 6 MeV. In the recent 20 years, elastic and inelastic neutron scattering cross sections were measured with high precision for a variety of elements at the PTB time-of-flight spectrometer. The D(d,n) reaction was primarily used for the production of neutrons. Because of the Q value of the reaction and the available deuteron energies, neutrons in the energy range from 6 MeV to 16 MeV can be produced. For the cross section measurement at lower energies, however, another neutron producing reaction is required. The {sup 15}N(p,n){sup 15}O reaction was chosen, as it allows the production of monoenergetic neutrons with up to 5.7MeV energy. In this work, the {sup 15}N(p,n) reaction was studied with focus on the suitability as a source for monoenergetic neutrons in scattering experiments. This includes the measurement of differential cross sections for the neutron producing reaction and the choice of optimum target conditions. Differential elastic and inelastic neutron scattering cross sections were measured for lead at four energies in the region from 2 MeV to 4 MeV incident neutron energy using the time-of-flight technique. A lead sample with natural isotopic composition was used. NE213 liquid scintillation detectors with well-known detection efficiencies were used for the detection of the scattered neutrons. Angle-integrated cross sections were determined by a Legendre polynomial expansion using least-squares methods. Additionally, measurements were carried out for isotopically pure {sup 209}Bi and {sup 181}Ta samples at 4 MeV incident neutron energy

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Mirror reactor studies

International Nuclear Information System (INIS)

Moir, R.W.; Barr, W.L.; Bender, D.J.

1976-01-01

Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80 percent. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59 percent and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high recirculating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)

Observation of Energy and Baseline Dependent Reactor Antineutrino Disappearance in the RENO Experiment.

Science.gov (United States)

Choi, J H; Choi, W Q; Choi, Y; Jang, H I; Jang, J S; Jeon, E J; Joo, K K; Kim, B R; Kim, H S; Kim, J Y; Kim, S B; Kim, S Y; Kim, W; Kim, Y D; Ko, Y; Lee, D H; Lim, I T; Pac, M Y; Park, I G; Park, J S; Park, R G; Seo, H; Seo, S H; Seon, Y G; Shin, C D; Siyeon, K; Yang, J H; Yeo, I S; Yu, I

2016-05-27

The RENO experiment has analyzed about 500 live days of data to observe an energy dependent disappearance of reactor ν[over ¯]_{e} by comparing their prompt signal spectra measured in two identical near and far detectors. In the period between August of 2011 and January of 2013, the far (near) detector observed 31 541 (290 775) electron antineutrino candidate events with a background fraction of 4.9% (2.8%). The measured prompt spectra show an excess of reactor ν[over ¯]_{e} around 5 MeV relative to the prediction from a most commonly used model. A clear energy and baseline dependent disappearance of reactor ν[over ¯]_{e} is observed in the deficit of the observed number of ν[over ¯]_{e}. Based on the measured far-to-near ratio of prompt spectra, we obtain sin^{2}2θ_{13}=0.082±0.009(stat)±0.006(syst) and |Δm_{ee}^{2}|=[2.62_{-0.23}^{+0.21}(stat)_{-0.13}^{+0.12}(syst)]×10^{-3}  eV^{2}.

A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

NARCIS (Netherlands)

Gallucci, F.; Basile, A.

2007-01-01

In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

Cronos 2: a neutronic simulation software for reactor core calculations

International Nuclear Information System (INIS)

Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

1999-01-01

The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

Siloe, Osiris, and the future perspective of swimming-pool reactors

International Nuclear Information System (INIS)

Chatoux, J.; Denielou, G.; Lerouge, B.

1964-01-01

Siloe and Osiris are two new general purpose research reactors of the 'Commissariat a l'energie Atomique'. Siloe, located within the 'Centre d'Etudes Nucleaires' of Grenoble is a swimming pool reactor of the same type as Melusine and Triton. It operates, at a nominal power of 15 MW thermal and has reached the peak power of 20 MW thermal with two thirds of its cooling system working. The fast flux above 1 MeV, which is maximum at the center of the core at 15 MW thermal is 1,2. 10 14 . The core, quite open, is downward cooled. Average specific power is 159 kW/l. Osiris is under construction at Saclay. Designed for 50 MW thermal, this reactor is upward cooled. The fast flux at the center of the core above 1 MeV is calculated to be 2, 5.10 14 . The average designed specific power is 280 kW/l. A fixed zircaloy gamma shield makes a box round the core. Future perspectives open to non-pressurised swimming-pool reactors are examined. Ways are suggested for neutronic; thermal and shielding modifications which make possible further improvements in the performances and economy of these devices. (authors) [fr

Experimental study of the passive flooding system in the WWER-1000 reactor

International Nuclear Information System (INIS)

Malyshev, A.B.; Efanov, A.D.; Kalyakin, S.G.

2002-01-01

The design solution of the passive flooding system in the WWER-1000 reactor core with the V-392 reactor facility and the scheme of the GE-2 large-scale thermohydraulic stand for substantiation of its functions are presented. The proposals, improving the efficiency of the system are developed on the basis of the experimental studies on the equipment input-output operational characteristics and the recommendations on the substantiation of the function of the reactor core flooding system are given [ru

Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

Energy Technology Data Exchange (ETDEWEB)

Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

2009-07-01

The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-12-31

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

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

A helium-cooled blanket design of the low aspect ratio reactor

International Nuclear Information System (INIS)

Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

1998-03-01

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh