Alternatives to reach safeguards goals at Atucha I nuclear power plant

International Nuclear Information System (INIS)

Palacios, E.; Orpet, P.; Marzo, M.; Valentino, L.; Vicens, H.

2001-01-01

Full text: This paper describes the main features of Atucha I Nuclear Power Plant and the current safeguards' approach applied to this installation by the International Atomic Energy Agency (IAEA) and the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials (ABACC). The reasons for not completely fulfilling the IAEA safeguards criteria with the current approach are also described and a conceptual proposal of an unattended system developed jointly by ABACC and the Nuclear Regulatory Authority of Argentina (ARN) is presented. Finally, the paper addresses an alternative proposal to the previous one aiming at fulfilling the above mentioned objectives. Atucha I Nuclear Power Plant (NPP) was built in the 70's and has been under operation since 1974, This is an On Load Reactor, moderated and refrigerated with heavy water (PHWR). From its starting up to about a year ago, this NPP operated with natural uranium fuel assemblies but presently the reactor core is fed with slightly enriched uranium fuel assemblies (0,85 %). This Plant generates up to 357 Mwe. An outstanding operating characteristic of this power reactor is that low burn-up fuels assemblies already discharged into the pond may be re-used when necessary upon neutron flux requirements (re-shuffling). This installation has a pond storage capacity of about 10,000 fuel assemblies. At the highest power rate, the reactor core must be fed with a frequency of about 0,72 fuel assemblies per day. Before the application of the Agency Safeguards Criteria (IAEA-SC) in (1991), Atucha l had always satisfied the IAEA safeguards goals. Since 1991 the IAEA-SC demanded for On Load Reactors the control of the flow of irradiated fuel assemblies that leave or enter into the core (re-shuffling). By that time, Atucha I had been working for about seventeen years and there was no possibilities to install specific safeguards equipment without making significant construction modifications on this installation. Under the

Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

Energy Technology Data Exchange (ETDEWEB)

Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I

1996-07-01

In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

Xenon oscillation in a large PHWR core (Atucha II type): TRISIC code applicability

International Nuclear Information System (INIS)

Solanilla, Roberto

2000-01-01

A three dimensional nuclear reactor simulation code (TRISIC) was developed many years ago to design a PHWR (pressurizer heavy water reactors - Atucha type) based in the 'source-sink model' (heterogeneous theory). The limited processor computational performance available at that time was the constraint of the code when a detailed reactor description was necessary. A modern PC (pentium) code version with a full reactor core representation (461 fuel channels) including diagonal control rod banks and flux-reading detectors with theirs tube guide was used in the present paper for simulation of the Xenon transient when a local asymmetric perturbation was produced in a large core (Atucha II type). The results obtained and the computer time required for the 70 hour's simulation with an adequate time step, established the potential of the code to deal with this kind of transients. The paper shows that the method of TRISIC allows to detect and control azimuthal, radial and axial oscillation. This code is a proper way to elaborate a program of control rods movement from the flux reading detectors to damp the oscillation. TRISIC could also be a accurate tool to supervise the full core flux distribution in real time during the operation of the reactor. (author)

Blackout sequence modeling for Atucha-I with MARCH3 code

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

A Coupled Calculation Suite for Atucha II Operational Transients Analysis

International Nuclear Information System (INIS)

Mazzantini, O.; Schivo, M.; Cesare, J.D.; Garbero, R.; Rivero, M.; Theler, G.

2011-01-01

While more than a decade ago reactor and thermal hydraulic calculations were tedious and often needed a lot of approximations and simplifications that forced the designers to take a very conservative approach, computational resources available nowadays allow engineers to cope with increasingly complex problems in a reasonable time. The use of best-estimate calculations provides tools to justify convenient engineering margins, reduces costs, and maximises economic benefits. In this direction, a suite of coupled best-estimate specific calculation codes was developed to analyse the behaviour of the Atucha II nuclear power plant in Argentina. The developed tool includes three-dimensional spatial neutron kinetics, a channel-level model of the core thermal hydraulics with subcooled boiling correlations, a one-dimensional model of the primary and secondary circuits including pumps, steam generators, heat exchangers, and the turbine with all their associated control loops, and a complete simulation of the reactor control, limitation, and protection system working in closed-loop conditions as a faithful representation of the real power plant. In the present paper, a description of the coupling scheme between the codes involved is given, and some examples of their application to Atucha II are shown

Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

International Nuclear Information System (INIS)

Baron, J.; Caballero, C.; Zarate, S.M.

1997-01-01

The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

The text of the agreement between the Agency and Argentina for the application of safeguards to the Atucha power reactor facility

International Nuclear Information System (INIS)

1995-01-01

The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Agreement of 3 October 1972 between Argentina and the IAEA for the application of safeguards to the Atucha Power Reactor Facility has been suspended

Integrated management for aging of Atucha Nuclear Power Plant

International Nuclear Information System (INIS)

Ranalli, J.M.; Marchena, M.H.; Sabransky, M.; Fonseca, M.; Santich, J.; Pedernera, P.

2012-01-01

Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina.. With the start-up of Atucha II and aiming to integrate the Ageing Management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) created an Ageing Management Department to cope with all ageing issues of both Atucha I and II. In this project both organization has formed a joint working group. The role of CNEA is providing technical support to the plant in the development of procedures a methodological framework for the Ageing Management Program of Atucha NPP. The main documents that have being issued so far are: . An Ageing Management Manual, including standard definition of Materials, Ageing Related Degradation Mechanisms, Operation Environments customized for Atucha NPP. . Walk down procedures and checklists aimed to systematize data collection during outages. . Procedures for performing Ageing Management Reviews and Maintenance Reviews for passive and active components. . Condition Assessments of several safety related systems. . Condition assessment of electrical components. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented (author)

Thermoelastic analysis for the fuel claddings of the nuclear power reactor at Atucha in the skid's region

International Nuclear Information System (INIS)

Sanchez Sarmiento, Gustavo; Basombrio, F.G.

1979-01-01

For the fuel elements of the Nuclear Power Reactor at Atucha, a two-dimensional thermoelastic analysis has been made in the region of the skids of the fuel cladding, when the gap between them and the fuel rod separator's support becomes zero. In such a case the latter forces exert on the skids an elastic reaction opposite to the cladding's expansion. The internal pressure reaching the yield stress for the cladding material has been calculated, as a function of the initial gap; for several possible fuel rod locations within the separator; for the actual dimensions and also for reduced thickness of the cladding; with a given external pressure and, with a known temperature spatial distribution. The latter has been calculated by solving the heat conduction equation along the fuel element for a certain power level in the reactor. The calculations are made with two FORTRAN IV computer codes developed at C.A.B., using the finite-element method: the NOLICUARM, to solve the nonlinear quasi-harmonic equation, and the ELASTEF 3, for the solution of thermoelastic problems with plane symmetry. (author) [es

Experimental and numerical determination of the dynamic properties of the reactor building of Atucha II NPP

International Nuclear Information System (INIS)

Ceballos, M.A.; Car, E.J.; Prato, T.A.; Prato, C.A.; Alvarez, L.M.; Godoy, A.R.

1995-01-01

Determination of the dynamic properties of the reactor building of Atucha II NPP is carried out in order to: i) Obtain valuable information for seismic qualification of the plant, and ii) Assess some procedures for testing and analysis that are used in the process of seismic evaluation of existing nuclear facilities founded on Quaternary soil deposits. Both steady state and impulsive dynamic tests were performed but attention is centered here in tile techniques used to determine natural frequencies and modal damping ratios with impulsive tests. Numerical analyses were performed by means of a 3-D model model of the superstructure together with foundation stiffness coefficients derived in a separate paper from steady state vibration tests, and also from analysis with a 2-D F.E. model of the soil layers capable of approximating the 3-D features of the problem. The computed foundation stiffness coefficients are compared both with those obtained from the tests and from an axisymmetric F.E. model; results indicate that foundation stiffness coefficients calculated with F.E. models with soil parameters given by laboratory tests performed on cored samples are significantly lower than those given by the steady state vibration tests. (author)

The pressure and leak tests in Atucha II

International Nuclear Information System (INIS)

Anon.

1991-01-01

This work deals with the pressure and leak tests of the containment sphere in the Atucha II Nuclear Power Plant's reactor building. This sphere is a metallic container, made in highly resistant steel plate, that is, built for providing the plant with a biological and structural barrier, which -in turn- provides safety and environmental protection. The applicable rules for these tests establish that the containment erection must be complete and in equivalent conditions to those that will prevail during the NPP operation. Particularly, pressure tests were carried out for assessing the structural condition of the sphere, while the leak test is aimed at the detection of tentative leaks [es

Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

International Nuclear Information System (INIS)

Chiaraviglio, N.; Bazzana, S.

2013-01-01

Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

Hydrogen combustion study in the containment of Atucha-I nuclear power plant

International Nuclear Information System (INIS)

Baron, J.H.; Gonzalez Videla, E.

1997-01-01

In this paper the combustion of hydrogen was modeled and studied in the containment vessel of the Atucha I nuclear power station using the CONTAIN package. The hydrogen comes from the oxidation of metallic materials during the severe accidents proposed. The CONTAIN package is an integrated tool that analyzes the physical, chemical and radiation conditions that affect the containment structure of the radioactive materials unloaded from the primary system during a severe accident in the reactor. (author) [es

Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant

International Nuclear Information System (INIS)

Lerner, Ana M.; Madariaga, Marcelo R.

1999-01-01

In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)

Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

International Nuclear Information System (INIS)

Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

1987-01-01

In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

Comparison of vibration test results for Atucha II NPP and large scale concrete block models

International Nuclear Information System (INIS)

Iizuka, S.; Konno, T.; Prato, C.A.

2001-01-01

In order to study the soil structure interaction of reactor building that could be constructed on a Quaternary soil, a comparison study of the soil structure interaction springs was performed between full scale vibration test results of Atucha II NPP and vibration test results of large scale concrete block models constructed on Quaternary soil. This comparison study provides a case data of soil structure interaction springs on Quaternary soil with different foundation size and stiffness. (author)

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

ATUCHA I NPP - Emergency drill practice

International Nuclear Information System (INIS)

Sanda, Alejandro; Rosales, Gabriel

2008-01-01

Full text: Atucha I NPP performs an Emergency Drill Practice once a year. Its main goals are: -) Fulfill the requirements of the Argentine Nuclear Regulatory Authority (ARN) regarding Atucha I NPP's Operating License; -) Fulfill the commitment with the community regarding the safe and reliable operation Atucha I NPP; -) Verify the response of the Civil Organizations, Security Forces, and Armed Forces, as well as the correct application of the Emergency Plan; -) Perform the 'General Alarm Drill' periodic control; -) Perform a re-training of the members of the Security Advisor Internal Committee (CIAS) on the Internal and External Aspects of the Emergency Plan and on the related procedures; -) Test the Emergency Communications System. New goals are added every year, considering the Drill's scope. This drill comprises two different kinds of practices: Internal practices (practices in the station, with our personnel) and external practices (practices outside the station with governmental organizations). Internal practices comprise: -) Internal and external communications practices; -) Acoustic alarms; -) Personnel gathering in the Meeting Points; -) Safety of selected Meeting Points; -) Personnel count, selective evacuation; -) Iodide Potassium pills distribution; -) CICE (Internal Group for Emergency Control) Coordination. External practices comprise: -) Nuclear Regulatory Authority; -) Argentine Navy, Comando Area Naval Fluvial, Base Naval Zarate; -) Lima firemen; -) Zarate firemen; -) Municipal Civil Defense (Zarate and Lima); -) National Guard, Escuadron Atucha; -) Zarate Regional Hospital; -) Lima Police Department; -) Zarate Police Department; -) Argentine Coast Guard, Zarate; -) Local radios: Radio FM Libre, FM El Sitio; -) First Aid clinic. The following activities are performed together with the aforementioned organizations: -) Formation of an 'Operative committee'; -) Evacuation of citizens in a 3 km radio; -) Control of every access to Lima; -) Control of

Water chemistry of Atucha II PHWVR. Design concepts and evolution

International Nuclear Information System (INIS)

Chocron, Mauricio; Rodriguez, Ivanna; Duca, Jorge; Fernandez, Ricardo; Rico, Jorge

2007-01-01

Full text: Atucha II is a pressurized heavy water vessel reactor designed by Siemens-KWU, currently part of AREVA NP, of 745 MWe and similar to Atucha I, which has been in operation over 25 years. The primary heat transport system (PHTS) is composed by vertical channels (277-313 C degrees) that allocate the fuel elements while the moderator circuit is composed by a partially separated circuit (142-173 C degrees). The moderation power is transferred to the feedwater through the moderator heat exchangers (HX). These HXs operate as the last, high pressure water-steam cycle heaters as well. Materials (with exception of fuel channels and fuel sheaths which are made of zirconium alloys) are all austenitic steels while cobalt containing alloys have been all replaced at the design stage. Steam generator and moderator HX tubing are Alloy 800 made. The core is operated without boron except with the first fresh nucleus. The secondary circuit or Balance of plant (BOP) is similar in conception to that of a PWR but the moderator HXs. It is entirely built of ferrous alloys, has a feedwater-deaerator tank and moisture separator. The energy sink is the Rio de la Plata River. The Reactors Chemistry Department, Chemistry Division, National Atomic Energy Commission, in its character of R and D institution has been committed by CNA II-N.A.S.A Project to prepare the water chemistry specifications, water chemistry engineering and manuals, considering the type of reactor, design and construction aspects and operation characteristics, taking into account the current state-of-the art and worldwide standards. This includes conceptual aspects and implementation and operative aspects as well. This documentation will be released after a designer's review as it has been stated in the respective agreement. Respecting the confidentiality agreement between CNEA and NASA and the confidentiality regarding handling original documentation provided by the designer, it is considered illustrative to

Fission distribution measurements of Atucha's fuel pellets with solid state track detectors

International Nuclear Information System (INIS)

Ricabarra, M.D. Bovisio de; Waisman, Dina.

1979-08-01

Distribution of fissions in a UO 2 rod has been measured by means of solid state detectors. Mica muscovite and Makrofol-N detectors were used in the experiment. The merits of mica muscovite relative to the Makrofol-N for the detection of fission fragments have been verified. However both fission track detectors closely agree (0,5%) in the final fission distribution of the UO 2 rod. Sensitivity of the detectors shows to be linear in the range between 50.000and 360.000 fission tracks per square centimeter. Due to the high spatial resolution this method is better than any other technique. Determination were made in UO 2 pellets similar to the fuel element of the Atucha reactor. The average fission rate in the rod has been measured within 0,8% error, and provides an accurate determination for the distribution of fissions in the rod wich is needed for the determination of energy liberated per fission in the natural uranium rod.(author) [es

The application of the PARCS neutronics code to the Atucha-I and Atucha-II NPPs

Energy Technology Data Exchange (ETDEWEB)

Ward, Andrew; Collins, Ben; Xu, Yunlin; Downar, Thomas [Purdue University, West Lafayette, IN (United States); Madariaga, Marcelo [Autoridad Nuclear Regulatoria, Buenos Aires (Argentina)

2008-07-01

In order to analyze Central Nuclear Atucha II (CNA-II) with coupled RELAP5/PARCS, extensive benchmarking of the neutronics codes HELIOS and PARCS was completed. This benchmarking was performed using a range of test problems designed in collaboration with NA-SA. HELIOS has been previously used to model Candu systems, but the results were validated for this case as well. The validation of both HELIOS and PARCS was performed primarily by comparisons to MCNP results for the same problems. Though originally designed to model light water systems, the capability of the PARCS was validated for predicting the performance of a Pressurized Heavy Water Reactor. The other noteworthy issue was the control rods. Because the insertion of the rods is oblique, a special routine was added to PARCS to treat this effect. Lattice level and Core level calculations were compared to the corresponding NA-SA codes WIMS and PUMA. In all cases there was good agreement in the results which provided confidence that the neutronics methods and the core neutronics modelling would not be a significant source of error in coupled RELAP5/PARCS calculations. (authors)

Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT

International Nuclear Information System (INIS)

Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto

2011-01-01

In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

Main pumps lost incident in the nuclear power plant Atucha I. Modelling with RELAP5/MOD3.2

International Nuclear Information System (INIS)

Ventura, M.A.; Rosso, R.D.

1998-01-01

Time evolution of natural circulation in the nuclear power plant Atucha I (CNA-I), in a main pumps lost incident because of the lost of external power feed, is analyzed. It leads to a strong stop transient, without an important blow down, from a forced nominal flow to a natural circulation one. The results are obtained from RELAP5/MOD3.2 code's modeling. The study is based on the refrigeration conditions analysis, during the first minutes of the reactor out of service. Previously to the transient, work had been done to obtain the plant steady state, with design parameters in operation conditions at 100 % of power. The object is that the actual plant state would be represented. In this way, each plant part (steam generators, reactor, pressurizer, pumps) had been modeled in separated form with the appropriate boundary conditions, to be used in the whole circuit simulation. The developed model, had been validated making use of the comparison between the values obtained to the principal thermodynamic parameters with the plant recorded values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/MOD3.2 capability to model the thermodynamic phenomena in a PHWR plant has been verified when, according to the mentioned incident, the flow pass from a nominal forced flow, to one which is governed by natural circulation, still with the CNA-I untypical design conditions. (author) [es

New fuel advanced heavy water reactors

International Nuclear Information System (INIS)

Notari, Carla

1999-01-01

A redesign of the PHWR fuel element (FE) to be used in all Argentine nuclear power plants has been proposed elsewhere. This new FE presents several characteristics aimed to an improved in-core performance and economical benefits derived from the unification of most of the fabrication processes that today constitute two different production lines: one for Embalse nuclear power plant CANDU type fuel and another for Atucha I. Atucha I and Embalse, the two operating nuclear power plants in Argentina, are PHWR of different conception. Atucha I (357 M we) is of pressure vessel type and the fuel elements are full-length assemblies (530 cm of active length) with 36 uranium rods in the cluster and a support one in the outer ring. Embalse (648 M we) is a CANDU pressure tube reactor fuelled with the well known 37 rod / 50 cm length fuel bundles, twelve of which are loaded in each channel. The more relevant changes in the proposed design are an increased subdivision of the fuel material in 52 rods and a 100 cm long bundle. The combined features give the adequate channel pressure drop. The proposed CARA design shows a superior neutronic performance than the standard PHWR fuel elements currently used in Atucha I and Embalse nuclear power plants. A variant of the CARA FE consisting in the elimination of the central four rods, leaving 48 rods and a central free space, is strongly recommended because it saves materials (less uranium, less sheaths) with no loss of burnup. The central D 2 O zone allows a better utilization of the inner rods and compensates the diminished uranium loading. In Embalse no differences in core physics are expected except the beneficial decrease in linear power density. In Atucha I besides the lower power density, a higher exit burnup appears as a consequence of the higher uranium inventory. The exit burnup figures have been calculated with cell and reactor models and the result is that similar fuel management schemes as the proposed for Atucha I for the

Atucha II nuclear power plant digital simulation

International Nuclear Information System (INIS)

Santome, D.; Rovere, L.A.T.

1987-01-01

This paper describes the start-up of a digital simulation code apt to be performed in real time of Atucha II nuclear power plant, foreseeing its subsequent usage in a Basic Principles Simulator. Adaptability and modification of existing routines and development of modules in order to incorporate the necessary variables dynamics to couple the different modes, were the main tasks. The mathematical model used allows the representation of the following sub-systems: a) a reactor's core point model, which comprehends the neutronic kinetics, fission and decaying powers, thermal transfer and Xe-poisoning calculation; b) pressurizer, which considers two sub-systems that may or may not be in thermodynamic equilibrium, both in two phases; c) coolants and moderators bonds considering separate moderator loops with the aim of introducing asymmetric perturbations; d) secondary sub-subsystem, which includes the feed water loop, pumps, steam generators and control valves; e) steam generators; f) control and safety systems, including power control, steam generators levels, moderator's temperature primary loop system, limitations and protection. (Author)

Full scale dynamic tests of Atucha II NPP

International Nuclear Information System (INIS)

Konno, T.; Alvarez, L.M.; Ceballos, M.A.; Prato, C.A.; Uchiyama, S.; Godoy, A.R.

1995-01-01

This paper summarizes the main results of a series of dynamic tests of the reactor building of Atucha II NPP performed to determine the dynamic properties of its massive structure deeply embedded in Quaternary soil deposits. Tests were performed under two different types of loading conditions: Steady state harmonic loads imposed by mechanical exciters and impulsive loads induced by dropping a weight on the ground surface in the vicinity. Natural frequencies and mode shapes were identified and the associated modal damping ratios were experimentally determined. Numerical analyses of the reactor building-foundation system by two different F.E. models were performed. One of them, based on an axisymmetric representation of the soil-structure system, was used to simulate the steady state vibration tests and to calculate the dynamic stiffness of the foundation slab and soil layers for comparison with those experimentally obtained. The other, a 3-D F.E. model of the superstructure, was used to assess the natural frequencies and mode shapes obtained from the tests, representing dynamic stiffness of the foundation with stiffness coefficients derived both from the tests and from the axisymmetric F.E. model. Good agreement of the natural frequencies given by two types of tests was generally found, with the largest difference between them in the fundamental frequency of the building. Estimates of modal damping derived from the tests showed significant differences depending on the technique used to calculate them. For the fundamental mode damping was found to be 23-42 %, gradually decreasing with frequency to 2-4 % for around 10 Hz. (author)

Full scale dynamic tests of Atucha II NPP

International Nuclear Information System (INIS)

Prato, C.A.; Ceballos, M.A.; Konno, T.; Uchiyama, S.; Alvarez, L.M.; Godoy, A.R.

1998-01-01

This paper summarizes the main results of a series of dynamic tests of the reactor building of Atucha II NPP performed to determine the dynamic properties of its massive structure deeply embedded in quaternary soil deposits. Tests were performed under two different types of loading conditions: steady state harmonic loads imposed by mechanical exciters and impulsive loads induced by dropping a weight on the ground surface in the vicinity. Natural frequencies and mode shapes were identified and the associated modal damping ratios were experimentally determined. Numerical analyses of the reactor building-foundation system by two different F.E. models were performed. One of them, based on an axisymmetric representation of the soil-structure system, was used to simulate the steady state vibration tests and to calculate the dynamic stiffness of the foundation slab and soil layers for comparison with those experimentally obtained. The other, a 3-D F.E. model of the superstructure, was used to assess the natural frequencies and mode shapes obtained from the tests, representing dynamic stiffness of the foundation with stiffness coefficients derived both from the tests and from the axisymmetric F.E. model. Good agreement of the natural frequencies given by two types of tests were generally found, with the largest difference between them in the fundamental frequency of the building. Estimates of modal damping derived from the tests showed significant differences depending on the technique used to calculate them. For the fundamental mode, damping was found to be 23-42%, gradually decreasing with frequency to 2-4% for ∝10 Hz. (orig.)

ND online software development for data acquisition of replacement operations of fuel assemblies of Atucha I Nuclear power plant

International Nuclear Information System (INIS)

Calvo, Maria Dolores; Wentzeis, Luis

2012-01-01

The ND Online software was developed in order to acquire data on a real-time basis of the refueling operations at the Atucha I nuclear power plant. The fuel elements containing slightly enriched uranium dioxide are located in the nuclear reactor core inside the cooling channels. The refueling operations are made periodically while the reactor is operating at full power. The acquired signals during the refueling operations are: pressure, force and position of the fuel element. In order to improve the safety and availability of the installation, monitoring of the refueling operations is important for the early detection of anomalies related to the fuel element itself, the cooling channels or the refueling machine (author)

Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

Energy Technology Data Exchange (ETDEWEB)

Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

2011-07-01

In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

Construction and cost experience regarding the 2nd pool house for spent fuel storage facility in the Atucha Power Station

International Nuclear Information System (INIS)

Barbosa, C.A.

1980-01-01

The Atucha I second pool house storage for spent fuel is designed as an extension of the Atucha I power station. The two are linked by civil structure, controlling circuits, electrical and compressed air and water supplies, low level wastes disposal, ventilation under pressure maintenance, and, most important, the ability to transfer spent and new fuel in both directions. Because the second pool house is, by location and design, an extension of the existing installation, and since there is no design departure, regarding storage and transfer of fuel from that of the original installation, the rules and regulations applied for its construction were the same as those valid for the Atucha I construction. The requirement not to exceed a four-year period for construction and commissioning was determined by the need to have storage room for the Atucha I fuel. Argentina will meet the 1982 target by having the installation available during the second half of 1981. The second pool house is a wet storage location with a capacity of 1000 tons metallic uranium. It was designed by the Kraftwerk Union of West Germany along the same lines as the 440-ton storage location originally built with the station. The Atomic Energy Commission of Argentina has managed the construction and participated in project and design stages. As in the original pool, the 6 m long assemblies are stacked in double tiers. The cost figures which are mentioned differ from previously released figures and are not the final ones. With civil construction almost finished and mechanical erection started, the present estimates should not differ by more than 10% from the final figures. The installation has an investment cost of 61 million dollars, (1980), and, depending on the amortization time span considered, a total yearly cost per kg of capacity of metallic uranium, ranging between 5.5 and 9.3 dollars per kg

Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

International Nuclear Information System (INIS)

Sidelnik, J.I.; Sosa, M.A.

1987-01-01

The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO 2 required. (Author)

Atucha I nuclear power plant surveillance programme

International Nuclear Information System (INIS)

Jinchuk, D.

1993-01-01

After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs

Atucha I nuclear power plant surveillance programme

Energy Technology Data Exchange (ETDEWEB)

Jinchuk, D [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1994-12-31

After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

Energy Technology Data Exchange (ETDEWEB)

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Peixoto, Orpet J.M. [Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials, Avenida Rio Branco, 123/Grupo 515- Centro, CEP: 20040-005, Rio de Janeiro (Brazil); Diaz, Gustavo [National Regulatory Authority - Argentina, Av. Del Libertador 8250, (1429) Buenos Aires (Argentina)

2015-07-01

uncertainties associated with both the calculations and measurements, a lower limit of the tolerance will be 15% with an upper limit of 20%. For the most part a 20% tolerance limit will be able to detect a missing assembly since in the vast majority of cases the drop in signal due to a single missing nearest neighbor assembly will be in the range 24-27%. The software was benchmarked against an extensive set of measured data taken at the site in 2004. Overall, 326 data points were examined and the prediction of the calibrated software was compared to the measurements within a set tolerance of ±20%. Of these, 283 of the predicted signals representing 87% of the total matched the measured data within ±10%. A further 27 or 8% were in the range of ±10-15% and 8 or 2.5% were in the range of ±15-20%. Thus, 97.5% of the data matched the measurements within the set tolerance limit of 20%, with 95% matching measured data with the lowest allowed tolerance limit of ±15%. The remaining 2.5% had measured signals that were very different from those at locations with very similar surrounding assemblies and the cause of these discrepancies could not be ascertained from the measurement logs. In summary, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will be deployed at the site and will enhance the capability of gross defect verification for the whole range of burnup, cooling time and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.

Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

International Nuclear Information System (INIS)

Ferraro, Diego

2008-01-01

Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

International Nuclear Information System (INIS)

Dyment, I.G.; Noguera Rojas, Francisco

1982-01-01

The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

Simulation program for the dynamic behaviour of the primary system and moderators's circuit of the Atucha I Nuclear Power Plant

International Nuclear Information System (INIS)

Castano, Jorge; Gvirtzman, H.A.

1981-01-01

A model of digital computation is presented to simulate the primary system of heat transportation, moderator system and the associated systems for adjustment, regulation and control in the PHWR reactor at the Atucha-1 nuclear power plant. The model discusses in a concentrated way the different components and allows the study of the dynamical behaviour of the power plant facing disturbances with respect to a state of stationary regime. General considerations and description of the model are made. The method is described showing flow sheets, graphs and developing basic formulas, simulating a primary system, moderator and secondary system of the steam generator and the main system of regulation. Also an analysis of the results is made, for the case of disturbances which reduce or increase the power of the reactor by 10%. (V.B.) [es

Upgrading Atucha 1 nuclear power plant. Regulatory perspective

International Nuclear Information System (INIS)

Caruso, G.

1998-01-01

Atucha 1 nuclear power plant has unique design and its commercial operation started in 1974. The upgrading decisions, the basis for an upgrading program and its status of implementation are presented. Regulatory decisions derived from the performance-based approach have the advantage that they enable balancing of the overall plant risk and identifying at different plant levels the areas where improvements are necessary. (author)

Atucha I nuclear power plant transients analysis

International Nuclear Information System (INIS)

Castano, J.; Schivo, M.

1987-01-01

A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

International Nuclear Information System (INIS)

Huber, Horacio

1989-01-01

Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

Atucha-I source terms for sequences initiated by transients

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

The present work is part of an expected source terms study in the Atucha I nuclear power plant during severe accidents. From the accident sequences with a significant probability to produce core damage, those initiated by operational transients have been identified as the most relevant. These sequences have some common characteristics, in the sense that all of them resume in the opening of the primary system safety valves, and leave this path open for the coolant loss. In the case these sequences continue as severe accidents, the same path will be used for the release of the radionuclides, from the core, through the primary system and to the containment. Later in the severe accident sequence, the failure of the pressure vessel will occur, and the corium will fall inside the reactor cavity, interacting with the concrete. During these processes, more radioactive products will be released inside the containment. In the present work the severe accident simulation initiated by a blackout is performed, from the point of view of the phenomenology of the behavior of the radioactive products, as they are transported in the piping, during the core-concrete interactions, and inside the containment buildings until it failure. The final result is the source term into the atmosphere. (author) [es

Modelling of blackout sequence at Atucha-1 using the MARCH3 code

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

Proposal of modification of the Atucha I nuclear power plant's emergency power supply system

International Nuclear Information System (INIS)

Palacio, Pedro; Dabove, Mario

1989-01-01

The emergency power supply system of Atucha I N.P.P. consists of three 50% diesel generators. During the transient from normal power supply to emergency power supply (approximately 15 seconds) an hydraulic generator takes care of the emergency system. By this way, the emergency busbars constitute themselves an interruption free system. The two emergency busbars work normally coupled. This proposal consists of the following modifications: 1) Add a new diesel generator in order to allow the operation with two diesel generators per busbar. 2) To work with the two emergency busbars not coupled as normal operation mode. 3) To eliminate the hydraulic generator from the emergency power supply system, in order to simplify the operation and to reduce the failure possibility. Without the hydraulic turbine generator, the emergency busbars loose the interruption free condition. For this reason, for the loads that are not able for this mode of operation and are connected to the emergency power supply system, two additional low-voltage interruption free busbars are necessary. Finally, this proposal is compared with the Atucha II N.P.P. emergency power supply system. (Author)

In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

International Nuclear Information System (INIS)

Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

2000-01-01

During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

Improvements related with the safety required by the Argentine Regulatory Authority to the Atucha I Nuclear Central; Mejoras relacionadas con la seguridad requeridas por la Autoridad Regulatoria Argentina a la Central Nuclear Atucha I

Energy Technology Data Exchange (ETDEWEB)

Calvo, J.; Michelin, C.; Navarro, R.; Waldman, R. [Autoridad Regulatoria Nuclear, Av. Del Libertador 8250, 1429- Ciudad de Buenos Aires (Argentina)]. e-mail: jcalvo@sede.arn.gov.ar

2006-07-01

The Argentinean Nuclear Regulation Authority (ARN) verified the existence of changes in the state of some internal components of the reactor of the Atucha I Nuclear Power station that, of continuing in the time, it could take to an inconvenient degradation for the safety operation of the installation. In consequence, to the effects of preventing that reach this situation, at the end of 1999, the ARN required to the Responsible Entity for the operation of this power station the implementation of an important improvements program in the internal components of the reactor. Additionally, and based on the results of the Probabilistic Safety analysis, it was added the one mentioned improvements program the implementation of an alternative cooling system of the reactor core denominated Second Drain of Heat, due to it was determined that, for some accidental sequences, their performance would reduce considerably the probability of damage to the core. The concretion of the improvements program implied to the Responsible Entity the realization of an important quantity of engineering studies, tests and specific inspections that allowed to carry out changes on the control bars of the reactor and its guide tubes; the coolant channels; the sensors of neutron flow; and diverse components of the primary and moderator systems. On the other hand also it was implemented the system Second Drain of Heat, what represents a considerable effort to make compatible the instrumentation and control of last generation, with the instrumentation and existent control systems in the power station. Also, it was requested to be carried out an integrity of the pressure recipient for to demonstrate the existence of an acceptable margin for the difference among the acceptable limit temperatures and of ductile/fragile transition of the material for all the possible accidental scenarios during the useful life of the reactor. (Author)

Model of automatic fuel management for the Atucha II nuclear central with the PUMA IV code

International Nuclear Information System (INIS)

Marconi G, J.F.; Tarazaga, A.E.; Romero, L.D.

2007-01-01

The Atucha II central is a heavy water power station and natural uranium. For this reason and due to the first floor reactivity excess that have this type of reactors, it is necessary to carry out a continuous fuel management and with the central in power (for the case of Atucha II every 0.7 days approximately). To maintain in operation these centrals and to achieve a good fuels economy, different types of negotiate of fuels that include areas and roads where the fuels displace inside the core are proved; it is necessary to prove the great majority of these managements in long periods in order to corroborate the behavior of the power station and the burnt of extraction of the fuel elements. To carry out this work it is of great help that a program implements the approaches to continue in each replacement, using the roads and areas of each administration type to prove, and this way to obtain as results the one regulations execution in the time and the average burnt of extraction of the fuel elements, being fundamental this last data for the operator company of the power station. To carry out the previous work it is necessary that a physicist with experience in fuel management proves each one of the possible managements, even those that quickly can be discarded if its don't fulfill with the regulatory standards or its possess an average extraction burnt too much low. For this it is of fundamental help that with an automatic model the different administrations are proven and lastly the physicist analyzes the more important cases. The pattern in question not only allows to program different types of roads and areas of fuel management, but rather it also foresees the possibility to disable some of the approaches. (Author)

Improvements related with the safety required by the Argentine Regulatory Authority to the Atucha I Nuclear Central

International Nuclear Information System (INIS)

Calvo, J.; Michelin, C.; Navarro, R.; Waldman, R.

2006-01-01

The Argentinean Nuclear Regulation Authority (ARN) verified the existence of changes in the state of some internal components of the reactor of the Atucha I Nuclear Power station that, of continuing in the time, it could take to an inconvenient degradation for the safety operation of the installation. In consequence, to the effects of preventing that reach this situation, at the end of 1999, the ARN required to the Responsible Entity for the operation of this power station the implementation of an important improvements program in the internal components of the reactor. Additionally, and based on the results of the Probabilistic Safety analysis, it was added the one mentioned improvements program the implementation of an alternative cooling system of the reactor core denominated Second Drain of Heat, due to it was determined that, for some accidental sequences, their performance would reduce considerably the probability of damage to the core. The concretion of the improvements program implied to the Responsible Entity the realization of an important quantity of engineering studies, tests and specific inspections that allowed to carry out changes on the control bars of the reactor and its guide tubes; the coolant channels; the sensors of neutron flow; and diverse components of the primary and moderator systems. On the other hand also it was implemented the system Second Drain of Heat, what represents a considerable effort to make compatible the instrumentation and control of last generation, with the instrumentation and existent control systems in the power station. Also, it was requested to be carried out an integrity of the pressure recipient for to demonstrate the existence of an acceptable margin for the difference among the acceptable limit temperatures and of ductile/fragile transition of the material for all the possible accidental scenarios during the useful life of the reactor. (Author)

Hydraulic Design of the CARA Fuel Assembly for Atucha-I

International Nuclear Information System (INIS)

Juanico, Luis; Brasnarof, Daniel

2000-01-01

In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

Operative modes of the primary circuit degasser of Atucha II N.P.P

International Nuclear Information System (INIS)

Rodriguez, Ivanna; Contino, Maximiliano; Chocron, Mauricio; Duca, Jorge

2012-09-01

Atucha II (N.A.S.A., Buenos Aires Province, Argentina) is a Pressurized Vessel Heavy Water Reactor designed by Siemens with a capacity of 740 MWe. After a long delay in construction the plant is close to the commissioning and among the many task that are carried out, chemistry and operation of devices related to it are under consideration [1]. As it is known, Hydrogen or Deuterium dosing has the purpose of both: limitation of the water radiolysis and to provide an appropriate reductive media for the structural materials, mainly stainless steel, A800 and Zr-4. Dealing with a heavy water plant, it is critical to determine whether it is necessary to add D 2 or if it is feasible to dose H 2 , by considering heavy water degradation and heavy water upgrading system capability. Those aspects have been previously analyzed and presented [2]. It is also necessary to consider blankets and venting locations that address to losses of the expensive D 2 . In the present work several alternatives of hydrogenation are presented and evaluated, considering the Degasser (D), the Volume Control Tank (TCV) and the special features of the purification and volume control system of a pressurized vessel heavy water plant where the primary circuit and moderator are partially mixed. Also the influence of venting through the pressurizer is analyzed. Conclusions are obtained in connection to (i) the maintenance of a permanent blanket of H 2 /He, 4%, in the TCV dome at a given initial pressure, (ii) The same but constant pressure to reach 0.6 ppm of H 2 in the Primary and Moderator water circuit, (iii) transients while reducing pressure in the Degasser and considering contribution of pressurizer venting, (iv) estimated contribution of the general corrosion of the system and (iv) differences if D 2 is used. (authors)

Overview of the SEU project for extended burnup at the Atucha-I NPP. Four years of operating experience

International Nuclear Information System (INIS)

Fink, J.M.; Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Casario, J.A.; Alvarez, L.

2002-01-01

Atucha I is a 357 MWe nuclear station moderated and cooled with heavy water, of German design located in Argentina. Fuelling is on-power and the plant was originally fuelled with natural uranium. To reduce fuel costs a program was initiated in August 1993 to introduce gradually slightly enriched uranium (SEU) fuel (0.85 w% U-235) with an associated burnup increase from 5900 MWd/tU to 11300 MWd/tU. The introduction of SEU fuel started in January 1995 and the program was divided in three Phases with an upper limit of SEU FA in the core: 12, 60 and 252 (full core) and licensing documentation was prepared for each Phase. This paper describes the most important aspects of the operating and project experience, and some factors limiting the burnup extension from an operation point of view. After four years of the program and with 181 SEU FA (71%) of the core, the operating experience has been good and without unfavourable effects due to the use of SEU fuel with the only exception of a small increase of the time to reach full power in plant startups or power cycling. In particular, the new criteria to prevent PCI failures in power ramps for higher burnup SEU fuel in refueling operations, plant startups or power cycling has been effective. The average discharge burnup of the SEU fuel taken out of the reactor in 1998 was 11263 MWd/tU. The average discharge burnup of the natural fuel in the same year was 6640 MWd/tU, with an increase of about 12% of the original value for a natural fuel core. The average number of fresh fuel assemblies per full power day was being reduced from 1.31 to 0.92 in 1998 and 0.83 in 1999. The fuel costs dropped gradually during the program from 9.38 (with natural uranium fuel) to 6.57 $/MWh in the first four months of 1999 (taking as reference the NU and SEU FA costs for 1999). Because of this the SEU program has been an important contribution to the reduction of Atucha I operating costs and to the competitiveness of nuclear power generation against

Atucha I nuclear power plant: repair works in QK02W01 moderator system heat exchanger; Central nuclar Atucha I. Intervencion al intercambiador nro2 del moderador

Energy Technology Data Exchange (ETDEWEB)

Olivieri, Luis E; Zanni, Pablo A [Nucleoelectrica Argentina SA (NASA), Lima (Argentina). Central Nuclear Atucha 1

2000-07-01

Atucha I nuclear power plant moderator system operates with highly radioactive heavy water, a pressure of 115 Bar and temperatures of about 200 C degrees. In March 2000, an increasing leakage of heavy water to the conventional thermal circuit was detected, conducting the plant to a shut down. The development of a number of actions and measures were taken, in order to plug this leakage. The leakage was found in a heat exchanger, which is located in a place of difficult access, with a high radiological yield and which, according to design, it was not considered to be mechanically repaired. It is a U bend tubes heat exchanger, weighting about 20 tons, and with a heavy water flow of 800 tons/h on the primary circuit, and 950 tons/h of ordinary water on the secondary side. Foreseeing this event, it had been designed and constructed special equipment and procedures, by means of a contract, with the Company INVAP SA. Repair works were carried out within a period of eighty-six (86) days, from which, forty five days were used to repair the component itself. A considerable amount of time was used to prepare simulators and the training of personnel. Due to the high radiological yield and the strict care of radiological standards, it was necessary the participation of 300 persons, integrating a collective dose of 4,86 Sv-m. It was necessary the construction of platforms and auxiliary stairs so as to make the work place accessible, as well as lifting and movement devices for heavy components, since this area does not have such kind of facilities. Welding and cutting machines remote controlled as well as manipulators which operated in front of the exchanger tube sheet were used. The aim was the reduction of dose values as much as possible. Special shielding were developed and in some cases it was necessary the adoption of drastic measures such as the cutting of bolts or pipes. The failure was detected and the tube was plugged. Also were plugged those tubes with wall thickness

Cost estimation of interim dry storage for Atucha I NPP

International Nuclear Information System (INIS)

Bergallo, Juan E.; Fuenzalida Troyano, Carlos S.

2007-01-01

A joint effort between NASA and CNEA has been performed in order to evaluate and fix the strategy of interim spent fuel storage for Atucha I nuclear power plant. In this work the cost estimation on the proposed system was performed in order to fix the parameter and design criteria for the next engineering step. The main results achieved show that both alternatives are all in the same range of costs per unit of mass to be stored, the impact on electricity cost is less than 1 US mills/KWh and the scaling factor achieved is 0.85. (author) [es

Micrometeorological study of the Atucha Nuclear Power Plant site

International Nuclear Information System (INIS)

Berri, G.J.; Robbio, C.A.

1986-01-01

The evaluation of time meteorological data obtained at the micrometeorological station of the Atucha Power Plant during 1979, is presented. Special attention is given to the transport and atmospheric dispersion characteristics through the evaluation of the mean and hourly wind behaviour and the stability classes. Furthermore, it is obtained an estimation of the dispersion factors both for short-term and long-term releases using Gaussians models. As these factors are representative of mean conditions, they should not be applied to the analysis of isolated situations. Finally it is emphasized that, although the results were obtained by means of 1979 data, significative differences are not expected for other years. (M.E.L.) [es

C A R A fuel element for Atucha nuclear power plants and development plan

International Nuclear Information System (INIS)

Brasnarof, D. O; Marino, A. C; Bianchi, D; Giorgis M A; Orlando, O; Munoz, C; Taboada, H; Florido, P. C

2006-01-01

This paper presents the current state and the development plan of the C A R A fuel element.Main activities were carried out towards to welding of the end plates of the C A R A fuel element by a new process, and the assembling and hanging of the C A R A fuel element in its Atucha configuration, by using an external basket [es

A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

Directory of Open Access Journals (Sweden)

Martina Adorni

2011-01-01

Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermalhydraulics, neutron physics, and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

The experience gained at various stages of the Atucha nuclear power plant project

International Nuclear Information System (INIS)

Cosentino, J.O.

1977-01-01

The paper describes the experience gained in Argentina at the successive stages of planning, feasibility study, decision-making, awarding of contracts, construction and operation of the first nuclear power plant in Latin America. In particular, the operating experience accumulated so far is summarized together with the requirements for preparing operating tables for the plant. The role of the Atucha plant is also described in connection with the second plant under construction and the third in the planning stage [es

Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

International Nuclear Information System (INIS)

1989-02-01

Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

Model and simulation of the hydraulic turbine speed regulator of the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Copparoni, G.; Etchepareborda, A.; Urrutia, G.

1992-01-01

The hydraulics turbines of Atucha I Nuclear Power Plant takes advantage of condenser cooling water level difference between the plant and the river to recover about 2,5 MW e. It also supplies emergency power until diesel generators start up. Speed regulation is needed due to the transients that during this process occur. The purpose is to minimize the diesels start up time, and to avoid overshoots on the internal grid frequency. The hydraulic turbine, its speed regulator and the electric system associated with this transient have been modeled. The models and some simulation results are presented in this work. (author)

Steam generator materials and secondary side water chemistry in nuclear power stations

International Nuclear Information System (INIS)

Rudelli, M.D.

1979-04-01

The main purpose of this work is to summarize the European and North American experiences regarding the materials used for the construction of the steam generators and their relative corrosion resistance considering the water chemestry control method. Reasons underlying decision for the adoption of Incoloy 800 as the material for the secondary steam generator system for Atucha I Nuclear Power Plant (Atucha Reactor) and Embalse de Rio III Nuclear Power Plant (Cordoba Reactor) are pointed out. Backup information taken into consideration for the decision of utilizing the All Volatil Treatment for the water chemistry control of the Cordoba Reactor is detailed. Also all the reasonswhich justify to continue with the congruent fosfatic method for the Atucha Reactor are analyzed. Some investigation objectives which would eventually permit the revision of the decisions taken on these subjects are proposed. (E.A.C.) [es

Regulatory actions towards dose reduction at Atucha 1 NPS

International Nuclear Information System (INIS)

Spano, F.; Curti, A.R.; Telleria, D.M.; Rudelli, M.D.

1998-01-01

Atucha 1, a nuclear power plant designed in the late sixties, is in commercial operation since June 1974. In some internal components such as the coolant channels, the station has Stellite-6, a high cobalt content alloy (up to 60%) for hard-facing application. The erosion and corrosion processes on the surfaces of the piping components of the primary coolant and moderator systems generate a varied type of particles oxides called 'crud'. The crud and cobalt 60 produced by neutron activation of cobalt are transferred by the water along the circuit of the coolant and moderator systems, producing deposits on internal surfaces. The cobalt deposits are dominant in radiation fields at working locations. For years, the Authority allowed a considerable number of station workers incurring doses near the limit since that installation had been built previously to introduction of the optimisation concept by ICRP publication 26. The recommendations included in ICRP publication 60 made more than difficult the radiological situation at the Atucha 1. For the facility, to comply with the new limit established by the Authority in January 1995, meant to carry out a substantial modification of the radiological conditions, specially the radiation fields due to cobalt 60. Some options to reduce individual and collective doses were analysed by the Authority. To carry out the evaluation of the deposit mechanisms and the real activity level of cobalt 60, a model of compartments connected by means of constant transfer coefficients was designed. It was concluded that there was a necessity to the change of coolant channels by new ones free of cobalt. It has been shown experimentally that radiation fields and occupational doses were reduced, due to the replacement programme carried out by the utility, in a similar way to the model predictions. At present after more than three years from the beginning of the application of the new limits, and after carrying out partially the tasks for the

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

De "átomos para la paz" a los reactores de potencia: Tecnología y política nuclear en la Argentina (1955-1976)

OpenAIRE

Hurtado de Mendoza, Diego

2005-01-01

Durante el período 1955-76, el programa nuclear argentino se integró a la arena internacional; su Comisión Nacional de Energía Atómica construyó cuatro reactores de investigación, adquirió a una empresa alemana y puso en marcha el primer reactor de potencia Atucha I, y compró a una empresa canadiense un segundo reactor de potencia. En este artículo se examinan estos desarrollos en relación con el contexto político local y con el panorama nuclear internacional. En particular, se analizan la po...

Activity determination for neutron dosimetry in the vigilance programme for the pressure vessel in Atucha I nuclear power plant

International Nuclear Information System (INIS)

Furnari, J.C.; Cohen, I.M.; Ciriani, D.F.; Helzel Garcia, J.

1993-01-01

The methodologies for the activity determination of Co-60, Nb-93m and Nb-94 in flux monitors are presented. This was done in order to evaluate dose and damage caused by radiation received by pressure vessel materials of the Atucha I nuclear power plant for its surveillance program. (author)

Experience in construction and operation of HWR plants in Argentina

International Nuclear Information System (INIS)

Madero, C.C.; Cosentino, J.O.

1982-01-01

''ATUCHA I'', the first nuclear power plant in Argentina, is in commerical operation since 1974 with a high capacity factor. The reactor is based on the MZFR prototype designed by SIEMENS with natural uranium and heavy water and PWR technic. The plant was built by SIEMENS on a turnkey contract and was rated 340 MWe. The offer presented in that opportunity by KWU was based on two reactors (ATUCHA I type) inside one single containment, due to the limitation in power of the reactor. Subsequent changes in the nature of the contract resulted in an active participation of CNEA engineering groups in the erection and commissioning of the reactor. In 1978 the national government approved a nuclear power plan to install four 600 MWe HWR plants until 1995. To start implementing this program, CNEA called for tenders for the supply of components and services for the ATUCHA II plant, in connection with the establishment of a local engineering company and the supply and construction of a heavy water production plant. In 1980 a contract was signed with KWU and the local company ENACE was formed to act as architect engineer and site coordinator. The plant will be located in the ATUCHA I site and the reactor will be similar but double in power to that one. Following the schedule of the nuclear plan, CNEA has just started preliminary studies for the next nuclear plant. ENACE will be responsible for the preparation of an offer for an ATUCHA reactor type. Local engineering and manufacturing firms, upon request and coordination from CNEA, and evaluating the local capacity to participate in the design and construction of a CANDU type nuclear plant. Final decision on this fourth nuclear plant in Argentina will be taken middle 1983. (J.P.N.)

Description of the PIE facility for research reactors irradiated fuels in CNEA

International Nuclear Information System (INIS)

Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

2002-01-01

The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

Mixed wastes treatment in Atucha I; Tratamiento de residuos mixtos de la CNA I (central nuclear Atucha I)

Energy Technology Data Exchange (ETDEWEB)

Varani, J L; Comandu, J F [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

1998-07-01

Full text: During decontamination works of the fueling machine of Atucha I nuclear power plant (AINPP), a liquid waste with special characteristics was generated, which needed the development of a treatment method. The waste consisted of an emulsion designed for the cleaning of mechanical components and was formed by an organic solvent dispersed in water with aid of an emulsifier additive. After several cleaning operations, the emulsion contained an important quantity of lubricants and radioactive dirt. The treatment had the objective of recycling a toxic waste such as the organic solvent and reducing the volume of the residual mass. Laboratory tests were made tending to the emulsion separation in their components. Ionic force and ionic mobility were modified for join the emulsion micelles and produce their coalescence. Different salts and working temperatures were tried and it was stated that the combination of 1% of Na{sub 2}SO{sub 4} added and 40 degree C temperature were the optimum taking into account the available equipment in AINPP and cost considerations. The process was carried out in batch mode and 3 residual streams were obtained, an aqueous one which was sent to Residual Water System of AINPP, an organic liquid consisting of decontaminated hydrocarbons, useful for other cleaning tasks and finally a solid one, sited in the in-between interface of the other two liquids, consisting of insoluble soaps used as lubricant thickness, containing the principal proportion of radioactivity. As a result of this process we have achieved a volume reduction higher than 90%, the recycling of the organic solvent and concentration of radioactivity in a solid greasy mass with low water solubility. (author) [Spanish] Texto completo: Como resultado de tareas de descontaminacion de la maquina de carga de la central nuclear Atucha I (CNAI) se genero un residuo liquido de caracteristicas especiales, que requirio el desarrollo de un metodo de tratamiento. El residuo consistia en

Mixed wastes treatment in Atucha I

International Nuclear Information System (INIS)

Varani, J.L.; Comandu, J.F.

1998-01-01

Full text: During decontamination works of the fueling machine of Atucha I nuclear power plant (AINPP), a liquid waste with special characteristics was generated, which needed the development of a treatment method. The waste consisted of an emulsion designed for the cleaning of mechanical components and was formed by an organic solvent dispersed in water with aid of an emulsifier additive. After several cleaning operations, the emulsion contained an important quantity of lubricants and radioactive dirt. The treatment had the objective of recycling a toxic waste such as the organic solvent and reducing the volume of the residual mass. Laboratory tests were made tending to the emulsion separation in their components. Ionic force and ionic mobility were modified for join the emulsion micelles and produce their coalescence. Different salts and working temperatures were tried and it was stated that the combination of 1% of Na 2 SO 4 added and 40 degree C temperature were the optimum taking into account the available equipment in AINPP and cost considerations. The process was carried out in batch mode and 3 residual streams were obtained, an aqueous one which was sent to Residual Water System of AINPP, an organic liquid consisting of decontaminated hydrocarbons, useful for other cleaning tasks and finally a solid one, sited in the in-between interface of the other two liquids, consisting of insoluble soaps used as lubricant thickness, containing the principal proportion of radioactivity. As a result of this process we have achieved a volume reduction higher than 90%, the recycling of the organic solvent and concentration of radioactivity in a solid greasy mass with low water solubility. (author) [es

Permission of change of limits in the vapor generators of the Atucha I Nuclear Central

International Nuclear Information System (INIS)

Ventura, M.

2006-01-01

In the mark of the modification of the Atucha-I Nuclear Central Installation (CNA-I) as consequence of the Introduction of the System 'Second Drain of Heat' (SSC), the Entity Responsible for the CNA-I (NASA) requested authorization to the Nuclear Regulatory Authority (ARN) to modify the value of the minimum level of water in the secondary side in the Steam generators (GVs) to activate the signal 'shoot of the Cut of the Reactor' (RESA-LLV). As the level in the GVs is one of those parameters that are used to shoot the Emergency Feeding System (RX), component of the SSC System, also was analyzed the change in the activation of the shoot signal of the 'Second Drain of Heat' (2SSC-LLV). The ARN uses for the study of the nuclear safety of nuclear power plants, the series of prediction programs RELAP5/MOD3.X. It participates of the evaluation and maintenance activities of these codes through specific agreements with the U.S. Nuclear Regulatory Commission (US-NRC). It is necessary to account with programs of this type since the ARN it licenses the construction and operation of Nuclear Power Plants (NPPs) and other outstanding facilities and it inquires its operation according to its own standards. With these tools its are auditing the calculations that the Responsible Entities of the operation make to guarantee the operability of the NPPs assisting the mentioned standards. The analysis with computational codes is used as a tool to achieve the best understanding in the behavior of the plant in union with the engineering approach, the manual calculations, the data analysis and the experience in the operation of the machine. (Author)

Measurement of the CNA I's (Atucha I nuclear power plant) control rods reactivity during its commissioning on January 8th, 1990

International Nuclear Information System (INIS)

Waldman, R.M.; Gomez, A.

1990-01-01

Measurements were made on integral and differential calibration of rod 16, fuel racks RG and R3 and extinction reactivity during Atucha I nuclear power plant's commissioning on January 8th., 1990. These were the first physical measurements performed after the first critical nuclear power plant's commissioning. (Author) [es

Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

International Nuclear Information System (INIS)

Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

1987-01-01

At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

International Nuclear Information System (INIS)

Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

1993-01-01

The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

Compact spent fuel storage at the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Antonaccio, Carlos; Conde, Alberto; Flores, Alexis; Masciotra, Humberto; Sala, Guillermo; Zanni, Pablo

2000-01-01

The object of this report is to verify the possibility to increase the available storage of irradiated fuel assemblies, placed in the spent fuel pools of the Atucha I nuclear power plant. There is intends the realization of structural modifications in the storage bracket-suspension beam (single and double) for the upper and lower level of the four spent fuel pools. With these modifications that increase the storage capacity 25%, would arrive until the year 2014, it dates dear for the limit of the commercial operation of nuclear power plant. The increase of the capacity in function of the permissible stress for the supports of the bracket-suspension beam. They should be carried out 5000 re-accommodations of irradiated fuel assemblies. The task would demand approximately 3 years. (author)

Nuclear Activities in Argentina. A short Review. Part 2

International Nuclear Information System (INIS)

Coll, Jorge A.; Radicella, Renato

2002-01-01

The second part of this historical review covers the 'industrial' period of nuclear energy in Argentina. The National Atomic Energy Commission (CNEA), after a feasibility study carried out by argentine experts, in 1968 signed a contract to build a nuclear power plant. This PHWR plant, Atucha 1, of 310 MWe was inaugurated in 1973 and it is still operating. The same year the CNEA signed a new contract to build a CANDU type plant of 600 MWe, Embalse, that was finally inaugurated in 1983. The construction of a third plant, Atucha 2 of 745 MWe also a PHWR, was started in 1980 but was arrested in 1994, when more than 80% was completed, and it is still waiting a political decision to reach completion. Within the development of the nuclear power program, a fuel element production plant for the Argentine power reactors was built by the CNEA and a heavy water production plant of 250 tons/year was inaugurated in 1993 in the southern province of Neuquen. A pilot spent fuel reprocessing plant was designed but its construction was not completed. At the same time, a pilot gaseous diffusion plant was constructed in order to produce enriched uranium for research reactors. The activities in the field of radioisotope and radiation applications were also intensive, mainly in nuclear medicine and food preservation. A facility to fabricate sealed sources was built to process the Co 60 produced by the Embalse power plant. Argentina was active in the export of nuclear facilities: CNEA built a complete nuclear research center in Peru, and the Argentine company INVAP built research reactors in Algeria and Egypt. The same company is now building a research reactor in Australia. (author)

Activities in Argentina related to the use of slightly enriched uranium in heavy water reactor NPPs

International Nuclear Information System (INIS)

Corcuera, R.

1999-01-01

An overview of activities related to the use of Slightly Enriched Uranium (SEU) fuel in HWR type NPPs, currently under execution in Argentina, is presented. The activities here described cover certain R and D lines as well as the main aspects of the Project 'Transition from full Natural-U to full SEU core in Atucha-I NPP'. Concerning the R and D lines, a summary is given on investigations related to reduction of void-coefficient using SEU fuel assemblies, annular pellet SEU fuel for bundle power flattening, etc. The main aspects of the above mentioned Project are outlined. At present, Atucha-I core is approaching a 40% of core load with SEU fuel, while the target of full SEU fuel core should be reached in 2-3 years. The expected exit burnup for such a core, namely 11000 MWD/tnU, is already currently obtained for SEU fuel in the present mixed core, while an increase in exit burnup of Natural-U fuel has also been obtained in very good agreement with reactor physics calculations. The comprehensive safety analysis carried out for each phase of the Project showed very moderated changes in plant behaviour under the set of postulated accidents and abnormal transients. A recent development, namely the CARA Project, aimed at unifying manufacturing of fuel assemblies for both operating NPPs in Argentina is presented in an accompanying paper. (author)

Concept and structure of instrumentation and control of the Atucha II nuclear power plant

International Nuclear Information System (INIS)

Garzon, D.; Roca, J.L.

1987-01-01

The general structure of instrumentation and control of Atucha II nuclear power plant as well as the technologies used, are described: concepts of functional decentralization and physical centralization; concept of functional group and functional complex; description of the technologies used (physical support) in the project of plant instrumentation and control; description of the different automation levels on the basis of concepts of control interface, automatism, regulation, group and subgroup controls; principles of signal conditioning; concept of announcement of alarms and state: supervisory computer, description of HAS (Hard wired Alarm System) and CAS (Computer Alarm System); application of the above mentioned structure to the project of another type of plants. (Author)

Implementation of the utilization program for the fuel elements of the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Martin, H.R.; Serra, O.H.; Parker, Alejandro

1981-01-01

The programming operation for the use of the fuel elements in the Atucha-1 nuclear power plant was initially under the responsibility of the KWU Company, as part of the services rendered due for the manufacturing of said elements. This job was done with the help of the TRISIC program, developed in the early seventies by CNEA and SIEMENS staff. From april 21, 1979 on, CNEA took over the responsibility and strategy of the interchange of fuel elements. The several stages carried out for the implementation of this service are detailed. (M.E.L.) [es

Growth scenarios with thorium fuel cycles in pressurised heavy water reactors

International Nuclear Information System (INIS)

Balakrishnan, M.R.

1991-01-01

Since India has generous deposits of thorium, the availability of thorium will not be a limiting factor in any growth scenario. It is fairly well accepted that the best system for utilisation of thorium is the heavy water reactor. The growth scenarios possible using thorium in HWRs are considered. The base has been taken as 50,000 tons of natural uranium and practically unlimited thorium. The reference reactor has been assumed to be the PHWR, and all other growth scenarios are compared with the growth scenario provided by the once-through natural cycle in the PHWR. Two reactor types have been considered: the heavy water moderated, heavy water cooled, pressure tube reactor, known as the PHWR; and the heavy water moderated and cooled pressure vessel kind, similar to the ATUCHA reactor in Argentina. For each reactor, a number of different fuel cycles have been studied. All these cycles have been based on thorium. These are: the self-sustaining equilibrium thorium cycle (SSET); the high conversion ratio high burnup cycle; and the once through thorium cycle (OTT). The cycle have been initiated in two ways: one is by starting the cycle with natural uranium, reprocessing the spent fuel to obtain plutonium, and use that plutonium to initiate the thorium cycle; the other is to enrich the uranium to about 2-3% U-235 (the so-called Low Enriched Uranium or LEU), and use the LEU to initiate the thorium cycle. Both cases have been studied, and growth scenarios have been projected for every one of the possible combinations. (author). 1 tab

Experimental and inspection facilities in post-irradiation of spent fuel pools for the analysis of the behaviour of nuclear fuels in power reactors

International Nuclear Information System (INIS)

Ruggirello, G.; Zawerucha, A.

1992-01-01

Since the beginning of the Atomic Nuclear Reactors (PHWR) Atucha I and Embalse in Argentine are employed different techniques for the knowing of the fuel bundles performances. It is detailed the facilities on post-irradiation examination. The techniques described are: online measurements, visual inspections, identifications of defective fuels and rods assemblies in spent fuel pools. This controls have made possible the feed-back to the manufactory process and the changes in the manufactory quality controls. (author)

Detection and analysis of thermal energy loss in the Atucha I nuclear power plant residual heat removal system

International Nuclear Information System (INIS)

Berra, Sandra; Guala, Mariana I.; Khon, Hector; Lorenzo, Andrea T.; Raffo Calderon, Maria C.; Urrutia, Guillermo

1999-01-01

It is presented the methodology used to detect and to measure energy losses which are existent in the Atucha I nuclear power plant. They were not directly detected, since the magnitude of those was below of the instrumentation precision which is used to measure the electric and thermal power in the plant. To achieve this work temperature special measurements were made. In this way it was possible to quantify the energy losses after operational long periods. (author)

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

Operational experience with the first eighteen slightly enriched uranium fuel assemblies in the Atucha-1 nuclear power plant

International Nuclear Information System (INIS)

Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Fink, J.; Casario, J.A.; Alvarez, L.

1997-01-01

Atucha I is a 357 Mwe nuclear station, moderated and cooled with heavy water, pressure vessel type of German design, located in Argentina. Fuel assemblies (FA) are 36 active natural UO2 rod clusters, 5.3 meters long and fuel is on power. Average FA exit burnup is 6 MWd/kg U. The reactor core contains 252 FA. To reduce the fuel costs about 6 MU$S/yr a program of utilization of SEU (0.85 %w U235) fuel was started at the beginning of 1995 with the introduction of 12 FA in the first step. The exit burnup of FA is approx. 10 MWd/kgU. It is planned to increase gradually the number of them up to having a full core with SEU fuel with an expected FA average exit burnup of 11 MWd/kgU. The SEU program has also the advantage of a strong reduction of spent fuel volume, and a moderate reduction of fuelling machine use. This paper presents the satisfactory operation experience with the introduction of the first 12 SEU fuel assemblies and the planned activities for the future. The fresh SEU fuel assemblies were introduced in six fuel channels located in an intermediate zone located 136 cm from the center of the reactor and selected because they have higher margins to the channel powers limits to accommodate the initial 15 to 20 % relative channel power increase. To verify the design and fuel management calculations, comparisons have been made of the calculated and measured values of the variation of channel ΔT, regulating rods insertion and flux reading in in-core detectors near to the refueled channel. The agreement was good and in most of the cases within the measurement errors. Cell calculations were made with WIMS-D4, and reactor calculations with PUMA. a fuel management 3D diffusion program developed in Argentina. With SEU fuel with a greater burnup in the central high power core region, new operating procedures were developed to prevent PCI failures in fuel power ramps that arise during operation. Some fuel rod and structural assembly design changes were introduced on the

Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Valle Cepero, Reinaldo

2004-01-01

The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

The Argentine nuclear policy: evaluation and proposals of the National Atomic Energy Commission. Privatization of the nuclear power plants and the situation of Atucha II. Addendum 1

International Nuclear Information System (INIS)

2001-01-01

An analysis is made of the possible privatization of the nuclear power plants in Argentina. The urgent definition of a nuclear policy for the short, medium and long term is strongly recommended to the Government together with the conclusion of the construction of the Atucha II nuclear power plant

Volatile amines treatment: Corrosion rates and Atucha I nuclear power plant experience

International Nuclear Information System (INIS)

Iglesias, Alberto M.; Jimenez Rebagliati, Raul; Raffo Calderon, Maria C.; Manzi, Ricardo

2000-01-01

Steam generators water treatment with volatile amines in place of ammonia is usual today. This option seems an acceptable alternative to the generalize use of ammonia-sodium phosphate and has advantages when copper alloys are present. There are several amines that can work as corrosion inhibitor but the most useful for plant applications are: morpholine, ethanolamine and cyclohexylamine. In this work, are present the obtained results of corrosion rates measurements by electrochemical methods. The hydrothermal conditions of our experiences were similar to that of the Atucha I nuclear power plant (CNA I). pH, conductivity and dissolved oxygen measures were correlated with corrosion rates of the CNA I materials as carbon steel and admiralty brass. The faradaic impedance spectroscopy techniques allows a more detailed interpretation of corrosion rates process. Morpholine and ammonia behavior can be evaluated under power plant operations conditions with the accumulated experience of CNA I. Results are present throughout material release and his effects over heat transfer parameters. (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

Development of an acoustic emission equipment for valves of the Nuclear Power Station Atucha 1

International Nuclear Information System (INIS)

Giaccheta, R.; Lopez Pumarega, I.; Straus, A.; Ruzzante, J.; Herzovich, P.

1994-01-01

A four channel Acoustic Emission was developed by the Acoustic Emission Group, INEND Department, of the Atomic Energy Commission of Argentina, for the detection of leaks in valves of the pressurized air system: ''Sistema de desconexion de emergencias por acido deuteroborico''. Basically, the system consists of four piezoelectric transducers with their corresponding preamplifiers coupled to the piping close to the valves. The following stages: amplifiers, threshold levels, channel identifications and visual alarm system are gathered in a box. The system was installed in the controlled zone of the Nuclear Power Stations Atucha I. It was calibrated and works on line. The values shown on the display are registered daily in order to separate the normal values from the leak ones. (author). 4 refs, 9 figs

IAEA’s Perspectives on Global Nuclear Power – Opportunities and Challenges

International Nuclear Information System (INIS)

Park, J.K.

2014-01-01

Status of global nuclear power: 437 reactors in operation (374.5 GWe); 2 reactors in long-term shutdown; 149 reactors in permanent shutdown; 70 reactors under construction. [As of Sep. 2014] Latest connections to the grid: - Ningde-2, 1000 MW(e), PWR, China; - Atucha-2, 692 MW(e), PHWR, Argentina; - Fuqing-1, 1000 MW(e), PWR, China). [Website: http://www.iaea.org/pris/]. IAEA projections of nuclear power: • Sep. 2014: 374.5 GWe; • 2030 - low 400.6 GWe: 7.0% increase; - high 699.2 GWe: 86.7% increase; • 2050 - low 412.9 GWe: 10.3% increase; - high 1091.7 GWe: 191.5% increase

Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

International Nuclear Information System (INIS)

Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

1990-01-01

This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

Spent fuel management of NPPs in Argentina

International Nuclear Information System (INIS)

Alvarez, D.E.; Lee Gonzalez, H.M.

2010-01-01

There are two Nuclear Power Plants in operation in Argentina: 'Atucha I' (unique PHWR design) in operation since 1974, and 'Embalse' (typical Candu reactor) which started operation in 1984. Both NPPs are operated by 'Nucleoelectrica Argentina S.A' which is responsible for the management and interim storage of spent fuel till the end of the operative life of the plants. A third NPP, 'Atucha II' is under construction, with a similar design of Atucha I. The legislative framework establishes that after final shutdown of a NPP the spent fuel will be transferred to the 'National Atomic Energy Commission', which is also responsible for the decommissioning of the Plants. In Atucha I, the spent fuel is stored underwater, until another option is implemented meanwhile in Embalse the spent fuel is stored during six years in pools and then it is moved to a dry storage. A decision about the fuel cycle back-end strategy will be taken before year 2030. (authors)

CARA Project: development of the advanced ULE fuel element for heavy water nuclear power plants

International Nuclear Information System (INIS)

Brasnarof, Daniel O.; Marino, Armando C.; Florido, Pablo C.; Munoz, C.; Bianchi, Daniel R.; Giorgis, Miguel A.

2006-01-01

The CARA Project (Spanish acronym of Combustible Avanzado para Reactores Argentinos) is a national fuel element technology development, compatible with our nuclear power plants (Atucha I, Embalse and Atucha II). It takes into account the experience obtained in our nuclear organisations (CNEA-CONUAR-NASA). The goal of the CARA fuel element is the performance improvement for those reactors and the enhancing of their normal operative conditions. The CARA design allows the burnup extension by using 52 rods of the same diameter. Likewise it keeps good thermo-hydraulic behaviour. The fuel bundle can be directly used in nuclear power plants with horizontal channels. By using an additional system it can be installed in the PHWR with vertical channels. The expected profits, by the use of the CARA in our reactors, broadly guaranty the recovery of the fund for its development, due to a reduction of the NPP fuels and back end cost. We estimate a reduction in the generation cost between 20 or 25 % in relation to the present one if we use 0.85 or 0.90% SEU (Slightly Enriched Uranium). The use of the CARA fuel in our reactors will also reduce the amount of spent fuel to be treated. The shortening could be between 17 to 27 % in Atucha I in relation to the present ULE (0.85%), between 38 to 46% for Embalse, and 45 to 53% for Atucha II. The mechanical behaviour and hydraulic compatibility have been verified. Several CARA prototypes were fabricated with a new design of the end plate and with new processes for the welding for the rods. We present in this paper the current status of the CARA fuel element development. (author) [es

Conceptual design of a quasi-homogeneous pressurized heavy water reactor to be operated in the closed Th-U233 fuel cycle

International Nuclear Information System (INIS)

1979-06-01

This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed

LDC nuclear power: Argentina

International Nuclear Information System (INIS)

Tweedale, D.L.

1982-01-01

Argentina's 31-year-old nuclear research and power program makes it a Third World leader and the preeminent Latin American country. Easily accessible uranium fuels the heavy water reactor, Atucha I, which provides 10% of the country's electric power. Atucha II and III are under construction. Several domestic and international factors combined to make Argentina's program succeed, but achieving fuel-cycle independence and the capacity to divert fissionable material to military uses is a cause for some concern. 60 references

Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

Energy Technology Data Exchange (ETDEWEB)

Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

2013-03-15

This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

Some conclusions obtained from the thermo-hydraulic behavior analysis of the nuclear power plant Atucha I, in case of loss of coolant accident with second heat sink

International Nuclear Information System (INIS)

Ventura, Mirta A.

2003-01-01

This paper is based on the recompilation, analysis and elaboration of the results of the operator (NA-SA), in the framework of the Atucha I Second Heat Sink project. The results have been compared with those obtained for the same power plant without second heat sink. The conclusions of the work permit the establishment of the operation rules of the plant. (author)

Conceptual design of interim storage facility for CNAI

International Nuclear Information System (INIS)

Fuenzalida Troyano, Carlos S.; Bergallo, Juan E.; Nassini, Horacio E.P.; Blanco, Anibal; Delmastro, Dario F.

2007-01-01

The reduced storage capacity available in the two spent fuel pools of argentine PHWR Atucha-1 power plant, the current plans for extending the reactor operation beyond its design lifetime, and the government decision on Atucha-2 NPP construction ending, have motivated the evaluation of a dry storage option for the interim management of spent fuel assemblies. Two different designs are presently being analyzed by an expert working group, from both technical and economical points of views. Authors are proposing a modular system consisting of an arrangement of reinforced concrete structures into which welded metallic canisters loaded with 37 spent fuel assemblies each stored in horizontal position. The reinforced concrete module is designed to provide the necessary physical protection and biological shielding to the loaded canisters during long-term storage, as well as passive means to remove the spent fuel decay heat by a combination of radiation, conduction and natural air convection. In this works are presented advances in the conceptual designs for a spent nuclear fuel system to Atucha I nuclear power plant. (author) [es

Hydrogeological modelling of the eastern region of Areco river locally detailed on Atucha I and II nuclear power plants area

International Nuclear Information System (INIS)

Grattone, Natalia I.; Fuentes, Nestor O.

2009-01-01

Water flow behaviour of Pampeano aquifer was modeled using Visual Mod-flow software Package 2.8.1 with the assumption of a free aquifer, within the region of the Areco river and extending to the rivers of 'Canada Honda' and 'de la Cruz'. Steady state regime was simulated and grid refinement allows obtaining locally detailed calculation in the area of Atucha I and II Nuclear power plants, in order to compute unsteady situations as the consequence of water flow variations from and to the aquifer, enabling the model to study the movement of possible contaminant particles in the hydrogeologic system. In this work the effects of rivers action, the recharge conditions and the flow lines are analyzed, taking always into account the range of reliability of obtained results, considering the incidence of uncertainties introduced by data input system, the estimates and interpolation of parameters used. (author)

Identification of limiting case between DBA and SBDBA (CL break area sensitivity): A new model for the boron injection system

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Gonzalez, R., E-mail: r.gonzalez@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Petruzzi, A., E-mail: a.petruzzi@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); D’Auria, F., E-mail: f.dauria@ing.unipi.it [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, 56122 San Piero a Grado, Pisa (Italy); Mazzantini, O., E-mail: mazzantini@na-sa.com.ar [Nucleo-electrica Argentina Sociedad Anonima (NA-SA), Buenos Aires (Argentina)

2014-08-15

Atucha-2 is a Siemens-designed Pressurized Heavy Water Reactor (PHWR) reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g. oblique Control Rods, Positive Void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank. The boron clouds reconstruction is obtained using a Computational Fluid Dynamics (CFD) CFX code calculation. A complete Large Break Loss Of Coolant Accident (LBLOCA) calculation implies the application of the RELAP5-3D{sup ©} system code. Within the framework of the third Agreement “Nucleoelèctrica Argentina-Sociedad Anonima (NA-SA) – University of Pisa/GRNSPG” (Contract, 2009), a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.

Design, test and start up of a cleaning system for the moderator tank bottom of Atucha I Nuclear Power Plant

International Nuclear Information System (INIS)

Duca, J.; Gerber, O.; Ibero, M.; Riga, N.

1989-01-01

In order to perform the cleaning of the moderator tank bottom, during the repair of the Atucha I nuclear power plant (CNA I) failure, the Empresa Nuclear Argentina de Centrales Electricas (ENACE S.A.) designed a system with the following requirements (asked by CNA I): a) To aspirate and retain free solid particles, uranium dioxide pellets and coolant channels isolations (foils) of minor size settled at the moderator tank bottom, being the reactor at middle loop state. b) To allow a radially cleaning up to 1.4 m from the extracted channel. c) To design a lay-out attaining the ALARA dose exposure. The designed system basically consists in: a) Flexible intake for suction: allows the movement inside the moderator tank and provides the adequate speed to raise the particles. b) Filter: retains the aspirated particles, pellets and foils. Its capacity is 1.8 dm 3 and the minimum size of retained particles is 200 m. The ALARA dose exposure concept is attained due to that the filter is located inside the moderator tank. c) Filtering column: contains the filter and allows the entrance of the extraction and exchange tool (for the flexible intake and filter). d) Suction hose: connects the filtering column with the pump. Its flexibility allows its use in any channel maintaining the same positions of the discharge pump and the return piping. e) Discharge pump: it is a canned centrifugal pump with low-low net positive suction head. f) Return piping: discharges the filtered water into the moderator tank. The system fulfilled satisfactorily all requirements during its operation. (Author)

Temporary storage facility for spent nuclear fuels at the Atucha I nuclear power station (CNA)

International Nuclear Information System (INIS)

Wasinger, K.

1983-01-01

According to plans of the Argentine Atomic Energy Commission (CNEA), the spent nuclear fuel elements of the Atucha I Nuclear Power Station are to be stored temporarily pending a decision about the ultimate disposal concept. The holding capacity of the first fuel storage facility built by the German KWU together with the whole power plant had been expanded in 1978 to a level good until mid-1982. In 1977, KWU drafted the concept of another fuel storage facility. Like the first one, it was designed as a wet storage system attached to the power plant installations and had a holding capacity of 6944 fuel elements, which corresponds to some 1100 te of uranium. This extends the storage capacity up until 1996. In 1978, KWU was commissioned by CNEA to plan the whole facility and deliver the mechanical and electrical equipment. CNEA themselves assumed responsibility for the construction work. The second fuel storage facility was commissioned three years after the start of construction. (orig.) [de

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

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

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.

Analysis of the main causes of failures in the Atucha I PWR moderator circuit branch piping

International Nuclear Information System (INIS)

Porto, J.; Sarmiento, G.S.

1983-01-01

From 1977 to 1979 four through cracks were detected in the auxiliary connection of the moderator piping with the coolant circuit in the PWR Atucha I Nuclear Plant. The failures were observed to occur systematically in the same place of the pipe, where mechanical stresses were detected experimentally and thermal stresses were calculated based on temperature values measured on the pipe. The temperature field in steady state conditions as well as during thermal shocks was modelled by finite element codes, and the corresponding thermal stresses were than numerically calculated. Considering those thermal and mechanical solicitations, a crack propagation analysis based on the elastoplastic fracture mechanics and the finite element method is now being developed. Among other causes such as fatigue corrosion and vibrations, the results of the analysis show that the most preponderant factors determining the cracking are mechanical stress, thermal stress and thermal fatigue

Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

1987-01-01

An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

Criticality and shielding calculations of an interim dry storage system for the spent fuel from Atucha I Nuclear Power Plant

International Nuclear Information System (INIS)

Silva, M

2006-01-01

The Atucha I Nuclear Power Plant (CNA-I) has enough room to store its spent fuel (SF) in damp in its two pool houses until the middle of 2015.Before that date there is the need to have an interim dry storage system for spent fuel that would make possible to empty at least one of the pools, whether to keep the plant operating if its useful life is extended, or to be able to empty the reactor core in case of decommissioning.Nucleolectrica Argentina S.A. (NA-SA) and the Comision Nacional de Energia Atomica (CNEA), due to their joint responsibility in the management of the SF, have proposed interim dry storage systems.These systems have to be evaluated in order to choose one of them by the end of 2006.In this work the Monte Carlo code MCNP was used to make the criticality and shielding calculations corresponding to the model proposed by CNEA.This model suggests the store of sealed containers with 36 or 37 SF in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.The results of the criticality calculations indicates that the solutions of SF proposed have widely fulfilled the requirements of subcriticality, even in supposed extreme accidental situations.Regarding the transference cask, the SF dose rate estimations allow us to make a feedback for the design aiming to the geometry and shielding improvements.Regarding the store modules, thicknesses ranges of concrete walls are suggested in order to fulfill the dose requirements stated by the Autoridad Regulatoria Nuclear Argentina [es

Commissioning of the laboratory of Atucha II NPP. Implementation and optimization of analytical techniques, quality aspects

International Nuclear Information System (INIS)

Schoenbrod, Betina; Quispe, Benjamin; Cattaneo, Alberto; Rodriguez, Ivanna; Chocron, Mauricio; Farias, Silvia

2012-09-01

Atucha II NPP is a Pressurized Vessel Heavy Water Reactor (PVHWR) of 740 MWe designed by SIEMENSKWU. After some years of delay, this NPP is in advanced construction state, being the beginning of commercial operation expected for 2013. Nucleoelectrica Argentina (N.A.S.A.) is the company in charge of the finalization of this project and the future operation of the plant. The Comision Nacional de Energia Atomica (C.N.E.A.) is the R and D nuclear institution in the country that, among many other topics, provides technical support to the stations. The Commissioning Chemistry Division of CNAII is in charge of the commissioning of the demineralization water plant and the organization of the chemical laboratory. The water plant started operating successfully in July 2010 and is providing the plant with nuclear grade purity water. Currently, in the conventional ('cold') laboratory several activities are taking place. On one hand, analytical techniques for the future operation of the plant are being tested and optimized. On the other hand, the laboratory is participating in the cleaning and conservation of the different components of the plant, providing technical support and the necessary analysis. To define the analytical techniques for the normal operation of the plant, the parameters to be measured and their range were established in the Chemistry Manual. The necessary equipment and reagents were bought. In this work, a summary of the analytical techniques that are being implemented and optimized is presented. Common anions (chloride, sulfate, fluoride, bromide and nitrate) are analyzed by ion chromatography. Cations, mainly sodium, are determined by absorption spectrometry. A UV-Vis spectrometer is used to determine silicates, iron, ammonia, DQO, total solids, true color and turbidity. TOC measurements are performed with a TOC analyzer. To optimize the methods, several parameters are evaluated: linearity, detection and quantification limits, precision and

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

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.

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

Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

International Nuclear Information System (INIS)

Kulichevsky, Raul M.

1995-01-01

Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

Ageing management program of wires for Atucha II Nuclear Power Plant

International Nuclear Information System (INIS)

Zorrilla, J.; Antonaccio, E.; Luraschi, C.; Simionato, E.; Nieto, H.

2012-01-01

Electrical cables constitute one of the most important components of NPP in terms o maintenance, safety and availability of the plant. Due to their large extension (thousands of kilometers) it is impossible to fully replace them and aging management becomes essential for long term operation of NPP. Since Atucha II is under construction there was a good opportunity to establish and implement a holistic cable aging management program from the beginning according to the state of the art in this field. The scope of this program involves safety related cables, including EQ cables and non EQ cables as well. Due to the diversity of the installed cables is impossible to address an aging management program of every single specimen. However it is possible to establish 'cables families' of similar aging behavior based on insulation and jacket material, manufacture, etc Several aging management strategies were set for the different 'cables families'. These strategies include cables deposit in plant, elaboration of procedures of visual and tactile inspection, NDE techniques, etc Currently the aging management program is being implemented covering topic such as: Cable screening and grouping. Review of EQ documentation and conventional qualification information referring to the installed cables. Establishment of base line of condition monitoring techniques. Calibration and set up of test parameters for NDE techniques. Aging mechanism characterization and determination of. Design construction and installation of cable deposit in plant (author)

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)

Atucha I nuclear power plant: repair works in QK02W01 moderator system heat exchanger

International Nuclear Information System (INIS)

Olivieri, Luis E.; Zanni, Pablo A.

2000-01-01

Atucha I nuclear power plant moderator system operates with highly radioactive heavy water, a pressure of 115 Bar and temperatures of about 200 C degrees. In March 2000, an increasing leakage of heavy water to the conventional thermal circuit was detected, conducting the plant to a shut down. The development of a number of actions and measures were taken, in order to plug this leakage. The leakage was found in a heat exchanger, which is located in a place of difficult access, with a high radiological yield and which, according to design, it was not considered to be mechanically repaired. It is a U bend tubes heat exchanger, weighting about 20 tons, and with a heavy water flow of 800 tons/h on the primary circuit, and 950 tons/h of ordinary water on the secondary side. Foreseeing this event, it had been designed and constructed special equipment and procedures, by means of a contract, with the Company INVAP SA. Repair works were carried out within a period of eighty-six (86) days, from which, forty five days were used to repair the component itself. A considerable amount of time was used to prepare simulators and the training of personnel. Due to the high radiological yield and the strict care of radiological standards, it was necessary the participation of 300 persons, integrating a collective dose of 4,86 Sv-m. It was necessary the construction of platforms and auxiliary stairs so as to make the work place accessible, as well as lifting and movement devices for heavy components, since this area does not have such kind of facilities. Welding and cutting machines remote controlled as well as manipulators which operated in front of the exchanger tube sheet were used. The aim was the reduction of dose values as much as possible. Special shielding were developed and in some cases it was necessary the adoption of drastic measures such as the cutting of bolts or pipes. The failure was detected and the tube was plugged. Also were plugged those tubes with wall thickness

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

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

Assessment of theoretical and experimental results in the calculation of atmospheric dilution factors in the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Quintana, E.E.; Tossi, M.H.; Telleria, D.M.

1990-01-01

Collective doses produced during the normal working of the Atucha I Nuclear Power Plant are calculated using annual atmospheric factors. This work studies the behaviour of the dilution factors in different periods of the year in order to fit the calculated dose model applying factors from seasonal, monthly or weekly periods. The Radiation Protection Group of the C.N.E.A. have carried out continuous environmental monitoring in the surroundings of the Atucha I Nuclear Power Plant. These studies include the measurement of air tritium concentration, radionuclide that is found principally as tritiated water vapour. This isotope, normally released by the nuclear power plant was used as a tracer to assess the atmospheric dilution factors. Factors were calculated by two methods: an experimental one, based on environmental measurements of the tritium concentration in the surroundings of the nuclear power plant and another one by applying a theoretical model based on information from the micrometeorological tower located in the mentioned place. To carry out the environmental monitoring, four monitoring stations in the surroundings of the power plant were chosen. Three of them are approximately one kilometer from the plant and the fourth is 7.5 km away, near the city of Lima. To condense and collect the atmospheric water vapour, an overcooling system was used. The measurement was performed by liquid scintillation counting, previous alkaline electrolytical enrichment of the samples. The theoretical model uses hourly values of direction and wind intensity, as well as the atmospheric dispersive properties. Values obtained during the period 1976 to 1988 allowed, applying statistical tests, to validate the theoretical model and to observe seasonal variation of the dilution factors throughout the same year and between different years. Finally, results and graphics are presented showing that the behaviour of the dilution factors in different periods of the year. It is recommended to

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

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

Design of an additional heat sink based on natural circulation in pressurized water reactors

International Nuclear Information System (INIS)

Frischengruber, Kurt; Solanilla, Roberto; Fernandez, Ricardo; Blumenkrantz, Arnaldo; Castano, Jorge

1989-01-01

Residual heat removal through the steam generators in Nuclear Power Plant with pressurized water reactors (PWR) or pressurized heavy water reactors (PHWR in pressured vessel or pressured tube types) requires the maintenance of the steam generator inventory and the availability of and appropriate heat sink, which are based on the operability of the steam generators feedwater system. This paper describes the conceptual design of an assured heat removal system which includes only passive elements and is based on natural circulation. The system can supplement the original systems of the plant. The new system includes a condenser/boiler heat exchanger to condense the steam produced in the steam generator, transferring the heat to the water of an open pool at atmospheric pressure. The condensed steam flows back to the steam generators by natural circulation effects. The performance of an Atucha type PHWR nuclear power station with and without the proposed system is calculated in an emergency power case for the first 5000 seconds after the incident. The analysis shows that the proposed system offers the possibility to cool-down the plant to a low energy state during several hours and avoids the repeated actuation of the primary and secondary system safety valves. (Author) [es

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

Permission of change of limits in the vapor generators of the Atucha I Nuclear Central; Permiso de cambio de limites en los GVs de la CNA-I

Energy Technology Data Exchange (ETDEWEB)

Ventura, M. [Autoridad Regulatoria Nuclear, Av. Libertador 8250 (1429), Capital Federal (Argentina)]. e-mail: mventura@sede.arn.gov.ar

2006-07-01

In the mark of the modification of the Atucha-I Nuclear Central Installation (CNA-I) as consequence of the Introduction of the System 'Second Drain of Heat' (SSC), the Entity Responsible for the CNA-I (NASA) requested authorization to the Nuclear Regulatory Authority (ARN) to modify the value of the minimum level of water in the secondary side in the Steam generators (GVs) to activate the signal 'shoot of the Cut of the Reactor' (RESA-LLV). As the level in the GVs is one of those parameters that are used to shoot the Emergency Feeding System (RX), component of the SSC System, also was analyzed the change in the activation of the shoot signal of the 'Second Drain of Heat' (2SSC-LLV). The ARN uses for the study of the nuclear safety of nuclear power plants, the series of prediction programs RELAP5/MOD3.X. It participates of the evaluation and maintenance activities of these codes through specific agreements with the U.S. Nuclear Regulatory Commission (US-NRC). It is necessary to account with programs of this type since the ARN it licenses the construction and operation of Nuclear Power Plants (NPPs) and other outstanding facilities and it inquires its operation according to its own standards. With these tools its are auditing the calculations that the Responsible Entities of the operation make to guarantee the operability of the NPPs assisting the mentioned standards. The analysis with computational codes is used as a tool to achieve the best understanding in the behavior of the plant in union with the engineering approach, the manual calculations, the data analysis and the experience in the operation of the machine. (Author)

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)

Argentina

International Nuclear Information System (INIS)

Tanis, S.; Ramberg, B.

1990-01-01

Argentina has embarked on an ambitious domestic nuclear research and energy program. The venture promises to provide the nation with the infrastructure to play an important role in the nuclear export market in the years ahead. Buenos Aires built its program on a foundation of international support and assistance that overlay an emerging sophisticated industrial and scientific establishment. The foreign contribution sensitized Argentina to opportunities in the global market. This paper reports that Argentina operates two heavy-water/natural uranium reactors, Atucha I and Embalse. Producing, respectively, 320 MW(e) and 600 MW(e), they supply roughly 10 percent of the country's electricity. With a capacity factor of 84 percent, Atucha I is among the best operated plants in the world. Plans call for a third, 745 MW(e) power plant, Atucha II, to go on line in the early 1990s. During this period, construction also may begin on three 300 MW(e) plants, although the country's foreign indebtedness plus cost overruns in earlier construction will place a heavy burden on these plans

Argentine activities within the IWGATWR framework

International Nuclear Information System (INIS)

Antunez, H.M.

1991-01-01

The status of power reactors and nuclear activity in Argentina for the period 1988-1990 are described. The reactor shutdown at the Central Nuclear Atucha I and repair works are discussed. Joint venture of the Comision Nacional de Energia Atomica and KWU developments are reported, including the conceptual design of an additional heat sink for pressurized water reactors, a feasibility analysis of the Co-60 production in pressure vessel PHWRs and a study of severe postulated accidents for pressure vessel PHWR. 11 refs, 4 figs

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

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)

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.

Experiments on ballooning in pressurized and transiently heated Zircaloy-4 tubes

International Nuclear Information System (INIS)

Markiewicz, M.E.; Erbacher, F.J.

1988-02-01

Single-rod burst tests were performed with Atucha I Zircaloy-4 cladding tubes in the REBEKA burst equipment of KfK. The objective was to investigate the ballooning and burst behavior of argentine cladding tubes obtained from NRG, Germany and CONVAR, Argentina. The burst data were compared with those of cladding tubes used in german PWR's. It was found that the burst data e.g. burst temperature, circumferential burst strain and its response to azimuthal temperature differences are identical for the Argentine and German tubing quality. The burst data are in good agreement with those of German PWR-Zircaloy tubes. Thus, the fuel rod behavior codes developed for German PWR's can also be used for the Argentine reactor Atucha I. (orig.) [de

Reassessment and suspension of the nuclear power plant design requirement of the constraint of collective dose per unit of practice. (Requirement 6 (b), Standard AR 3.1.2)

International Nuclear Information System (INIS)

Amado, Valeria A.; Canoba, Analia C.; Curti, Adriana R.; Biaggio, Alfredo L.

2009-01-01

By the middle of 2005, the Nuclear Regulatory Authority (ARN) decided to re-assess the basis of a design requirement applicable to the limitation of nuclear power reactor radioactive discharges. Such requirement, aimed at restricting the discharge of globally dispersed long-lived radionuclides, was in force in Argentina since 1979 and was expressed as a limitation of the collective dose commitment per unit of electrical energy generated. The practical result of such regulatory action was the need to retain C-14 in the Atucha II power reactor under construction as well as in future heavy water reactors to be built in the country, and, later on, to manage it as to assure its isolation from the biosphere during an appropriate period of time. For the above-mentioned reassessment, an ad hoc task group was created and an internal report was presented to the Board of Directors by the middle of 2007. Because of such report the ARN decided to suspend the application of the requirement (i.e. it is not more mandatory, even for Atucha II). The present work presents the main aspects of that report. In particular, it explains the basis of the design requirement and the most important assumptions that triggered it. The differences between the assumptions made at that time and the reality of nuclear power generation at the beginning of the 21st Century, as well as their implications in relation to the requirement are described, including the Suess effect and its impact in the total dose due to C-14. Finally, after explaining in detail the facts that made no longer reasonable to keep in force the above mentioned requirement, the work presents the conclusions that lead the ARN to the suspension of this requirement. (author) [es

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

Outlook for a new design application for the joint union geometry of zircaloy-4 pipe-plug welds

International Nuclear Information System (INIS)

Martinez, R.L; Corso, H.L; Ausas, J; Fernandez, L

2008-01-01

The potential advantages are described and the test results shown for a new joint design for the Zircaloy-4 sheath-plug welding, used in the production of fuel elements for nuclear reactor power generators. Samples taken from sheaths and plugs similar to those used in the Atucha I Reactor were welded using the orbital GTAW process (Gas Tungsten Arc Welding), using a new joint design and equipment that is now applied to the production of high quality welding, with very low levels of contamination. The test results from the experiments with hydrogen content, corrosion, metallographic and traction on samples of these welds are compared with those obtained in simples taken from the conventionally processed fuel elements for Atucha I. The results are also presented for the characterization of samples obtained with the same orbital welding method, but with a smaller protective chamber. This work aimed to verify the influence of the welding chamber size on the contamination with hydrogen. The utility of applying the new design together with the orbital GTAW method to the fabrication process for fuel elements is discussed

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.

Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I; Comparacion entre un modelo de diferencias finitas (PUMA) y uno de elementos finitos (DELFIN) para la simulacion del reactor de la CNA-I (central nuclear Atucha-I)

Energy Technology Data Exchange (ETDEWEB)

Grant, C R [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

1997-12-31

The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs.

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

37-Active rods fuel element for Atucha 1 nuclear power plant. Effects of this change in design over the neutronic behavior, decay power and radioactive inventory

International Nuclear Information System (INIS)

Villar, Javier E.

1999-01-01

The influence of the use of 37-rods fuel element on the behavior of the Atucha 1 nuclear power plant homogeneous core with slightly enriched fuel to 0.85 w % were studied through representative parameters such as average discharge burnup, channel powers, reactivity coefficients, kinetic parameters, radioactive inventory and decay power. In general, the values of mentioned parameters are similar to those corresponding to a core with the 36-rods fuel element actually in use, although it must be emphasized a decrease both in linear power and, in minor degree, in the efficiency of shut-off and control rods and a slight increase in the discharge burnup. The fuel management strategy developed for a core with 36-rods elements can be maintained. (author)

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)

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

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

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

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

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)

Nuclear Activities in Argentina, 2010

International Nuclear Information System (INIS)

Ferreri, J.C.; Ferreri, J.C.; Clausse, A.; Clausse, A.; Clausse, A.; Ordonez, J.P.; Mazzantini, O.A.

2011-01-01

Nuclear activities in Argentina are restarted. After almost two decades of near stagnation, the governments political decision of August 2006 regarding electrical energy production, considered the nuclear option as a valid one to solve the problems of the growing demand of electrical energy. This decision triggered again the activities related to the finalization of the third nuclear power reactor (Atucha-II), now actively progressing, the construction of a prototype of the CAREM integral advanced reactor, the life extension of the Embalse CANDU nuclear power plant (NPP) and the studies for the emplacement of a fourth NPP in an appropriate site. In all those years of near stagnation, there were notable exceptions related to the design and construction of experimental and radioisotope production reactors, led by INVAP, a state-owned industry, which exported its production. The accompanying industries of nuclear fuel elements production also remained active, given the demand of the two active NPPs. Meanwhile, the National Atomic Energy Commission of Argentina continued the efforts on research and development that were at the base of the technological achievements of the nuclear activities in Argentina. Nuclear safety studies associated with Atucha II and Embalse NPPs and radiological safety were also a substantive part of the continued efforts by Nucleo-Electrica de Argentina SA and the Nuclear Regulatory Authority of Argentina

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.

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

International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the Embalse CANDU reactor

International Nuclear Information System (INIS)

Rouben, B.; Chow, H.C.; Leung, L.K.H.; Inch, W.; Fink, J.; Moreno, C.

2004-01-01

In the last few years, Nucleoelectrica Argentina S.A. and Atomic Energy of Canada Limited have collaborated on a study of the technical feasibility of implementing Slightly Enriched Uranium (SEU) fuel in the Embalse CANDU reactor in Argentina. The successful conversion to SEU fuel of the other Argentine heavy-water reactor, Atucha 1, served as a good example. SEU presents an attractive incentive from the point of view of fuel utilization: if fuel enriched to 0.9% 235 U were used in Embalse instead of natural uranium, the average fuel discharge burnup would increase significantly (by a factor of about 2), with consequent reduction in fuel requirements, leading to lower fuel-cycle costs and a large reduction in spent-fuel volume per unit energy produced. Another advantage is the change in the axial power shape: with SEU fuel, the maximum bundle power in a channel decreases and shifts towards the coolant inlet end, consequently increasing the thermalhydraulics safety margin. Two SEU fuel carriers, the traditional 37-element bundle and the 43-element CANFLEX bundle, which has enhanced thermalhydraulic characteristics as well as lower peak linear element ratings, have been examined. The feasibility study gave the organizations an excellent opportunity to perform cooperatively a large number of analyses, e.g., in reactor physics, thermalhydraulics, fuel performance, and safety. A Draft Plan for a Demonstration Irradiation of SEU fuel in Embalse was prepared. Safety analyses have been performed for a number of hypothetical accidents, such as Large Loss of Coolant, Loss of Reactivity Control, and an off-normal condition corresponding to introducing 8 SEU bundles in a channel (instead of 2 or 4 bundles). There are concrete safety improvements which result from the reduced maximum bundle powers and their shift towards the inlet end of the fuel channel. Further improvements in safety margins would accrue with CANFLEX. In conclusion, the analyses identified no issues that

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)

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)

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

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)

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)

Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I

International Nuclear Information System (INIS)

Grant, C.R.

1996-01-01

The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs

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

Relation between the punctual defects dynamics and irradiation growth in Zr and its alloys

International Nuclear Information System (INIS)

Sarce, A.; Monti, A.M.; Smetniansky-De Grande, N.

1990-01-01

The knowledge of the static properties punctual defects and their interaction with the drains present in the material, such as dislocations and grain borders, is of fundamental importance for the correct prediction of the reactor's components behaviour under irradiation. This work calculates the Zr alloys growth under irradiation and results are compared with the stretching values of the cooling channels measured at the Atucha I nuclear power plant. (Author) [es

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

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

A challenging task: cleaning and repairing at nuclear power plant ATUCHA I (CAN-I) the primary's moderators cooling circuits heat exchangers

International Nuclear Information System (INIS)

Amaya, D.; Alaniz, A.; Bernasconi, R.

2005-01-01

A set of automatic and semiautomatic machines and tools were designed for the accomplishment of remotely controlled works in high radiation fields on the ATUCHA-I moderator heat exchangers. The object of this equipment is to carry out works related to the cleaning, inspection and eventual blocking of the heat exchanger's tubes. Due to the special characteristics of the area, such as difficult access, not much space and high dose rates, the remote operation of highly trained and specialized personnel with specially designed tools, is mandatory. The Principal operations consist of: 1. Equipment manually taken to the area by specialized personnel. 2. Remote cutting the bolting and cutting and re-weld seals with custom designed equipment. 3. Remote cutting and re-weld piping connections with equipment on customized tracks, special supports, drives and commands. 4. Remote cleaning, leak testing, machining and plugging of the tube sheets with a custom-made master-slave/Cartesian robotic manipulators. 5. Monitoring with video cameras and lighting systems incorporated into the equipment. 6. Ands others task as piping stabilization, supporting and moving flanges, re-alignment of seals and pipes, etc. This paper describes the entire development of this project, starting from the initial work plan to the completion of the first on-site work carried out at the facility. Including descriptions, drawings and pictures of the custom designed equipment, description of the performed works and comparisons between the actual doses and estimated manual operation doses. (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

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.).

PSA Level 2 as element of an integral safety assessment before plant commissioning

International Nuclear Information System (INIS)

Loeffler, H.; Mildenberger, O.; Sonnenkalb, M.; Steinroetter, T.

2012-01-01

In Argentina the Central Nuclear Atucha II is near to completion. This is a pressurized heavy water reactor. PSA (Probability Safety Assessment) level 1, level 2 and level 3 have to be performed in order to show compliance with the Argentinean dose limit. Such studies have been done first by the former KWU in the 1980's to get the construction license (FABIAN 1985). Nowadays the plant owner NA-SA performs PSA level 1 and provides information about the core damage states to GRS, who does the subsequent PSA level 2 part. GRS delivers source terms to the environment and the associated frequencies to the Argentinean research institute CNEA, which performs level 3 together with NA-SA. Since GRS is situated in the middle of the chain, interface definition with both ends has been a significant task of the GRS activities. Experience gained during this process will be highlighted in the presentation. The analysis of PSA level 2 proper follows a traditional approach: -) deterministic accident simulation with integral code MELCOR; -) analyses of specific issues which are not covered by MELCOR; and -) probabilistic accident progression analysis with EVNTRE event tree methodology. It appears that MELCOR and EVNTRE and PSA guidelines in general are flexible enough to analyse new or uncommon reactor designs. It also appears that the plant specific design features may require analyses beyond present code capabilities, calling for expert judgment and they can largely determine PSA results. The behaviour of iodine is not yet covered satisfactorily by state-of-the-art models in MELCOR

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

Chemical and radiochemical control of the primary circuit of Atucha INPP (Nuclear Power Plant) since the start up in January 1990

International Nuclear Information System (INIS)

Ali, S.P; Baungartner, E.C.; Blesa, M.A.

1990-01-01

Since the start up of Atucha I Nuclear Power Plant in January 1990, an exhaustive chemical and radiochemical control of primary media was undertaken. The main objectives were the evaluation of the water condition after the long outage and the determination of activity measurements limitations to detect and localize fuel failures. Chemical and radiochemical techniques were critically proved. At the same time, a complete program of updating and optimization of those procedures was developed, including the revision of the analytical parameters, range of applicability and accuracy. A more adequate processing of data was adopted. They were compared with historical values corresponding to periods with and without fuel elements failures, used as references. The analysis of theoretical models of total gamma activity concentration and some specific radionuclides activity concentration evolution and their rates, and the comparison with experimental data obtained during normal operation including some failure events, generated tables of alarm criteria through a combination of parameters. Additionally, actions are suggested for different combination of parameters. Operative conditions that might interfere in the detection and localization of a failed fuel element are also pointed out. (Author)

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)

Comparative analysis of C A R A fuel element in argentinean PHWR Argentinas

International Nuclear Information System (INIS)

Brasnarof D; Marino, A. C; Florido, P. C; Daverio, H

2006-01-01

This paper presents an analysis of the thermal mechanical behaviour, fuel consumption and economical estimations of the CARA fuel element in the Atucha and Embalse nuclear power plants, compared with the present fuel performance.The present results show that the expect profit by the use of the CARA fuel element in our reactor guaranties the recovery of fund for its development. Likewise it reduces the number of spent fuel to be storage and treated [es

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

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

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

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)

Sampling and characterization of spent exchange resins of Atucha I nuclear power plant

International Nuclear Information System (INIS)

Varani, Jose L.; Cernadas, D.; Iglesias, Alberto M.; Raffo Calderon, Maria del C.

2004-01-01

The present storage facilities for spent resins in Atucha I NPP would be full within the next 5 years, accordingly some tasks are being planned to conditioning these waste and so generate enough storage capacity for the remaining life of the plant. Among this tasks is the characterization of spent resins that has several objectives: to know their chemical and radiochemical composition; the homogeneity of these parameters in the total volume of spent resins; the existence or not of compact zones; the proportion of 'crud'; the breakage degree of the micro-spheres; etc. The first step was to analyse the criterion to follow for sampling resins in the storage deposit of 40 m 3 . In order to take some samples from different points, a special device was required. It had to be introduced closed in the resin bed, opened to take the sample and then closed again to return to the surface. A device used in cereal industry for sampling silos to different depths was modified in its internal capacity for reducing operator dose and increasing the length of rod in order to reaching the bottom of the pit. The device was tested in cold mock up before to taking actual samples. Active resins samples, five in total up to now, were taken from deposit to different depths and kept in lead containers. After analysing the samples, the following average results were extracted: 1.7 x 10 5 Bq/g of Co-60, 9.7 x 10 5 Bq/g of Cs-137 and 774 Bq/g of total alpha, which corresponds to intermediate activity waste. The differences between the values of activity of the different samples are of up to 310 % for Co-60 and of up to 788 % for the Cs-137 what indicates a great inhomogeneity. The direct observation of resin grains, placed in a transparent glass burette, did not demonstrate an important proportion of broken or divided resins. (author)

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)

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)

GIS Application in Atucha I Nuclear Power Plant Exercise Argentina, 2007

International Nuclear Information System (INIS)

Sadaniowski, I.V.; Telleria, D.M.; Jordan, O.D.; Boutet, L.I.; Kunst, J.J.; Bruno, H.A.; Hernandez, D.G.; Rodriguez, M.; Cateriano, M.A.; Rey, H.L.

2011-01-01

Geographic Information Systems (GIS) are tools applied to assist in the assessment and solution of many geographical related issues. Recently, their applications have been extended to the areas of disasters and environmental emergencies. GIS not only could be used as a diagnostic tool. Combined with adequate information and other tools capable to predict the transfer of pollutants in the environment and the associated impacts to the public, GIS could be used to support emergency planning and response. The Nuclear Regulatory Authority (NRA) of Argentina has incorporated in 2003 the GIS technology like an innovative resource for its preparation and response activities in emergencies. For this, the NRA acquired the necessary technology (hardware and software) and the technical specialist who were joined to expert's team in the nuclear and radiological emergencies field. The GIS stays operative as support tool in the Emergencies Control Center of NRA. In this paper, the use of GIS as a tool for analysis and advice is presented. The GIS is being used for preparation and development of nuclear emergencies trials and exercises, carried out on-site and off-site at the Nuclear Power Plant Atucha I Buenos Aires, Argentina, in cooperation with civil defense, national and state security and army forces and intensive public involvement. The databases were conformed with information from different sources, including the result of interviews to different actors, as well as other local and national government agencies and forces. Also, educational institutions, local medical centers, etc., were consulted. The information was enriched with outings to field in the surroundings of nuclear power plant. The scope and the detail of the information for this exercise covers 30 kilometers surroundings the nuclear power plant, with a range of significantly different geographical and population conditions. When loading the information in the GIS, a classification scheme is applied and

GIS application in Atucha I nuclear power plant exercise Argentina, 2007

International Nuclear Information System (INIS)

Sadaniowski, Ivana; Jordan, Osvaldo; Boutet, Luis; Kunst, Juan; Bruno, Hector; Hernandez, Daniel; Rodriguez, Monica; Cateriano, Miguel; Rey, Hugo; Telleria, Diego

2008-01-01

Full text: Geographic Information Systems (GIS) are tools applied to assist in the assessment and solution of many geographical related issues. Recently, their applications have been extended to the areas of disasters and environmental emergencies. GIS not only could be used as a diagnostic tool. Combined with adequate information and other tools capable to predict the transfer of pollutants in the environment and the associated impacts to the public, GIS could be used to support emergency planning and response. The Nuclear Regulatory Authority (NRA) of Argentina has incorporated in 2003 the GIS technology like an innovative resource for its preparation and response activities in emergencies. For this, the NRA acquired the necessary technology (hardware and software) and the technical specialist who were joined to expert's team in the nuclear and radiological emergencies field. The GIS stays operative as support tool in the Emergencies Control Center of NRA. In this paper, the use of GIS as a tool for analysis and advice is presented. The GIS is being used for preparation and development of nuclear emergencies trials and exercises, carried out on-site and off-site at the Nuclear Power Plant Atucha I Buenos Aires, Argentina, in cooperation with civil defense, national and state security and army forces and intensive public involvement. The databases were conformed with information from different sources, including the result of interviews to different actors, as well as other local and national government agencies and forces. Also, educational institutions, local medical centers, etc., were consulted. The information was enriched with outings to field in the surroundings of nuclear power plant. The scope and the detail of the information for this exercise covers 30 kilometers surroundings the nuclear power plant, with a range of significantly different geographical and population conditions. When loading the information in the GIS, a classification scheme is applied

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)

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

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)

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)

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

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

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

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.

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.

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)

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)

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.

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

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

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)

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)

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

Study of the Atucha I nuclear power plant's residual heat removal system unavailability through the fault tree analysis and common cause failures

International Nuclear Information System (INIS)

Terrado, C.A.

1991-06-01

The present essay offers a comprehensive research of the Atucha I nuclear power plant's residual heat removal system unavailability, including Fault Tree Analysis and Common Cause Failures (CCF) treatment. The study is developed within the Event Tree perspective that considers the loss of external electrical power of the initiating event. The event was constructed by the Safety Evaluations Division of the Ezeiza Atomic Center in Argentina. According to the Event Tree, the research includes system demand during plant operation with 132 KV and emergency generation (Diesel motor generators). The system unavailability assessment is approached in two different ways: a) Considering independent failures only. b) Taking into account the existence of Common Cause Events, and modeling dependent failures. The Fault Tree quantification is played using the AIEA PSAPACK Code. The assessment data base is compiled from plant specific records and generic data bases like TECDOC 478. After Fault Tree model logic development, some general procedures used in common cause failures treating are applied to pick up another set of solutions. The results of the study are: a) Four Fault Trees have been developed to model the abovementioned system: 132 KV and emergency generation, both including and excluding CCF. b) The following unavailability values were obtained: 132 KV independent failures only: 7 10 -4 . Emergency generation independent failures only: 1.53 10 -2 . 132 KV dependent and independent failures: 3.6 10 -3 . Emergency generation dependent and independent failures: 1.74 10 -2 . The major conclusions obtained from the precedent results are: a) When using 132 KV system configuration, minimal cut sets involving common cause failures represents 81%from total system unavailability. b) The dependent failures treatment is an important task to be considered in safety assessments in order to reach more realistic values. (Author) [es

Analysis of a PHWR slightly enriched fuel

International Nuclear Information System (INIS)

Notari, C.; Marajofsky, A.

1994-01-01

It is widely known that the use of slightly enriched uranium in PHWR reactors presents economic advantages derived from the fact that less uranium is required for producing the same amount of energy. Several studies related with the use of this alternative in Atucha I NPP have been performed. The fuel assembly geometry considered up to now has been almost identical to the natural uranium one. In this work a modification consisting in the use of annular pellets in the outer ring of the cluster is analyzed. This design produces several performance benefits. The redistribution of the power in the fuel improves the maximum to average bundle power ratio. The improvement achieved depends on the void volume in the pellets which at the same time represents a certain burnup decrease. These parameters (power ratios and burnup loss) are quantified for the Atucha I and Embalse NPPs. This design improves the fuel behaviour with respect to the burnup extension derived from the slight enrichment. It is also interesting in case an overall power increase is considered. (author). 16 refs, 8 figs, 1 tab

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)

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

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)

Set of rules SOR 2 reactor site criteria

International Nuclear Information System (INIS)

1976-06-01

The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-01-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

International Nuclear Information System (INIS)

Richard P. Wells

2007-01-01

This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-11-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

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

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

Assistance in chemistry and chemical processes related to primary, secondary and ancillary systems of nuclear power plants

International Nuclear Information System (INIS)

Chocron, Mauricio A.; Becquart, Elena T.; Iglesias, Alberto M.; La Gamma, Ana M.; Villegas, Marina

2003-01-01

Argentina is currently running two nuclear power plants: Atucha I (CNA I) and Embalse (CNE) operated by Nucleoelectrica Argentina (NASA) whereas the National Atomic Energy Commission (CNEA), among other activities, is responsible for research and development in the nuclear field, operates research reactors and carries out projects related to them. In particular, the Reactor Chemistry Section personnel (currently part of the Chemistry Dept.) has been working on the field of reactor water chemistry for more than 25 years, on research and support to the NPPs chemistry department. Though the most relevant tasks have been connected to primary and secondary circuits chemistry, ancillary systems show along the time unexpected problems or feasible improvements originated in the undergoing operating time as well as in phenomena not foreseen by the constructors. In the present paper are presented the tasks performed in relation to the following systems of Embalse NPP: 1) Heavy water upgrade column preliminary water treatment; 2) Liquid waste system preliminary water treatment; and 3) Primary heat transport system coolant crud composition. (author)

Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

International Nuclear Information System (INIS)

Abou Yehia, H.; Bars, G.; Tran Dai

2001-01-01

After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Apollo-L2, an advanced fuel tokamak reactor utilizing direct conversion

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Blanchard, J.P.; El-Guebaly, L.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.E.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Witt, R.J.

1989-01-01

A scoping study of a tokamak reactor fueled by a D- 3 He plasma is presented. The Apollo D- 3 He tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The low neutron wall loading (0.1 MW/m 2 ) permits a first wall lasting the life of the plant and enables the reactor to be classified as inherently safe. The cost of electricity is less than that from a similar power level DT reactor. 10 refs., 1 fig., 4 tabs

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Rapid data acquisition from the safety system of the FRJ-2 reactor

International Nuclear Information System (INIS)

Inhoven, H.

1980-06-01

The central department for research reactors (ZFR) of the Juelich Nuclear Research Centre (KFA) is operating the reactors FRJ-1 (MERLIN) and FRJ-2 (DIDO) since 1962. In 1976, a Siemens 330 computer has been put into operation especially for the processing of data from the DIDO reactor, followed by another computer of the same type for the purpose of processing data from the ZFR department in general. The present report is a result of the work investigating 'Data acquisition and data processing in the FRJ-2' and primarily discusses the complex of 'fast analog and binary signals'. The activities in this field of work have been and still are mainly concerned with general problems encountered in adapting a currently 14-year-old reactor system to a digital computer, namely problems such as data decoupling in the safety system of the reactor, data acquisition using the CAMAC system, data transfer via an 'extended branch', data acquisition software as core-resident programs, temporary storage as common data, interpreting software as peripheral - storage - resident programs. (orig./WB) [de

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

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

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

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

Estimation of power feedback parameters of pulse reactor IBR-2M on transients

International Nuclear Information System (INIS)

Pepyolyshev, Yu.N.; Popov, A.K.

2013-01-01

Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

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

PSA Level 2 activities for RBMK reactors

International Nuclear Information System (INIS)

Gubler, R.

1998-01-01

Probabilistic safety analyses (PSAs) of the boiling water graphite moderated pressure tube reactors (RBMKs) have been developed only recently and they are limited to Level 1. Activities at the IAEA were first motivated because of the difficulties to characterize core damage for RBMK reactors. Core damage probability is used in documents of the IAEA as a convenient single valued measure, for example for probabilistic safety criteria. The limited number of PSAs that have been completed for the RBMK reactors have shown that several special features of these channel type reactors necessitate revisiting of the characterization of core damage for these reactors. Furthermore, it has become increasingly evident that detailed deterministic analysis of DBAs and beyond design basis accidents reveal considerable insights into RBMK response to various accident conditions. These analyses can also help in better characterizing the outstanding phenomenological uncertainties, improved EOPs and AM strategies, including potential risk-beneficial accident negative backfits. The deterministic efforts should be focused first on elucidating accident progression processes and phenomena, and second on finding, qualifying and implementing procedures to minimize the risk of severe accident states The IAEA PSA procedures were mainly developed in New of vessel type LWRs, and would therefore require extensions to make them directly applicable. to channel type reactors. (author) (author)

Studsvik's R2 reactor - Review of activities

Energy Technology Data Exchange (ETDEWEB)

Grounes, Mikael; Tomani, Hans; Graeslund, Christian; Rundquist, Hans; Skoeld, Kurt [Studsvik Nuclear AB, Nykoeping (Sweden)

1993-07-01

A general description of the R2 reactor, its associated facilities and its history is given. The facilities and range of work are described for the following types of activities: fuel testing, materials testing, neutron transmutation doping of silicon, activation analysis, radioisotope production and basic research including thermal neutron scattering, nuclear chemistry and neutron capture radiography. (author)

Decommissioning of reactor facilities (2). Required technology

International Nuclear Information System (INIS)

Yanagihara, Satoshi

2014-01-01

Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

Reproduction of the PSBR reactor with Exterminator-2; Reproduccion del reactor PSBR con exterminador-2

Energy Technology Data Exchange (ETDEWEB)

Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1983-08-15

To reproduce the reactor PSBR reported in (1), with the available version of the Exterminator-II in the ININ, they took the dimensions, composition specifications, effective sections of the different compositions (excepting those of the central thimble and of the moderator), the K{sub eff} and the factors of power (FP) for the different burners. Based on the comparison of the K{sub eff} and of the FP obtained with those reported the precision it is determined before in the reproduction of the reactor mentioned. (Author)

Production of Sn-117m in the BR2 high-flux reactor.

Science.gov (United States)

Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

2009-01-01

The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

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

International Nuclear Information System (INIS)

1966-01-01

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

Comparative study between fluidized bed and fixed bed reactors in methane reforming with CO2 and O2 to produce syngas

International Nuclear Information System (INIS)

Jing Qiangshan; Lou Hui; Mo Liuye; Zheng Xiaoming

2006-01-01

Reforming of methane with carbon dioxide and oxygen was investigated over Ni/MgO-SiO 2 catalysts using fixed bed and fluidized bed reactors. The conversions of CH 4 and CO 2 in a fluidized bed reactor were close to thermodynamic equilibrium. The activity and stability of the catalyst in the fixed bed reactor were lower than that in the fluidized bed reactor due to carbon deposition and nickel sintering. TGA and TEM techniques were used to characterize the spent catalysts. The results showed that a lot of whisker carbon was found on the catalyst in the rear of the fixed bed reactor, and no deposited carbon was observed on the catalysts in the fluidized bed reactor after reaction. It is suggested that this phenomenon is related to a permanent circulation of catalyst particles between the oxygen rich and oxygen free zones. That is, fluidization of the catalysts in the fluidized bed reactor favors inhibiting deposited carbon and thermal uniformity in the reactor

Benchmark testing of Canadol-2.1 for heavy water reactor

International Nuclear Information System (INIS)

Liu Ping

1999-01-01

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

Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

International Nuclear Information System (INIS)

Burley, E.L.

1982-07-01

A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

Characterization of fuel distributions in the Three-Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

1984-04-01

The resolution of technical issues generated by the accident at Three-Mile Island Unit 2 (TMI-2) will inevitably be of long range benefit. Determination of the fuel debris dispersal in the TMI-2 reactor system represents a major technical issue. In reactor recovery operations, such as for the safe handling and final disposal of TMI-2 waste, quantitative fuel assessments are being conducted throughout the reactor core and primary coolant system

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

International Nuclear Information System (INIS)

Bissani, M; O'Kelly, D S

2006-01-01

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

Energy Technology Data Exchange (ETDEWEB)

Bissani, M; O' Kelly, D S

2006-05-08

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

International Nuclear Information System (INIS)

Elnokity, O.E.M.

2006-01-01

this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

General outline of the operation and utilization of the BR2 reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Leonard, F.; Gandolfo, J.M.; Lenders, H.

1978-01-01

The BR2 reactor is a high-flux material testing reactor of the thermal heterogeneous type. The fuel is 93% 235 U enriched uranium in the form of plates clad in aluminium. The moderator consists of beryllium and light water, the water being pressurized (12.5kg/cm 2 )and acting also as coolant. The pressure vessel is of aluminium, and is placed in a pool of demineralized water. One should stress the following main features of the design: the experimental channels are skew, the tube bundle presenting the form of a hyperboloid of revolution (see figure 1)-this gives easy access at the top and bottom reactor covers allowing complex instrumented devices, while maintaining a very high neutron flux at the core; great flexibilty of utilization, due to the fact that it is possible to adapt the core configuration to the experimental loading as the fissile charge can be centred on different experimental channels; although BR2 is a thermal reactor, it is possible to achieve neutron spectra very similar to those obtained in a fast reactor, either by the use of absorbing screens or by the use of fissile material within the experimental device; five 200mm diameter channels are available for loading large experimental irradiation devices, as in-pile sodium, gas or water loops. (author)

Benchmark testing of CENDL-2 for U-fuel thermal reactors

International Nuclear Information System (INIS)

Zhang Baocheng; Liu Guisheng; Liu Ping

1995-01-01

Based on CENDL-2, NJOY-WIMS code system was used to generate 69-group constants, and do benchmark testing for TRX-1,2; BAPL-UO-2-1,2,3; ZEEP-1,2,3. All the results proved that CENDL-2 is reliable for thermal reactor calculations. (3 tabs.)

Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant; Caracterizacion Mecanica y Radiologica de Partes de Canales Refrigerantes Irradiados Extraidos del Reactor de la Central Nuclear Atucha I

Energy Technology Data Exchange (ETDEWEB)

Piquin, Ruben [Instituto Balseiro, Universidad Nacional de Cuyo, Centro Atomico Bariloche, Universidad Nacional de Mar del Plata (Argentina)

2001-07-01

The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

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

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

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

The 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)

Digital, remote control system for a 2-MW research reactor

International Nuclear Information System (INIS)

Battle, R.E.; Corbett, G.K.

1988-01-01

A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

A nodal Grean's function method of reactor core fuel management code, NGCFM2D

International Nuclear Information System (INIS)

Li Dongsheng; Yao Dong.

1987-01-01

This paper presents the mathematical model and program structure of the nodal Green's function method of reactor core fuel management code, NGCFM2D. Computing results of some reactor cores by NGCFM2D are analysed and compared with other codes

RHTF 2, a 1200 MWe high temperature reactor

International Nuclear Information System (INIS)

Brisbois, Jacques

1978-01-01

After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

Turkey's regulatory plans for high enriched to low enriched conversion of TR-2 reactor core

International Nuclear Information System (INIS)

Guelol Oezdere, Oya

2003-01-01

Turkey is a developing country and has three nuclear facilities two of which are research reactors and one pilot fuel production plant. One of the two research reactors is TR-2 which is located in Cekmece site in Istanbul. TR-2 Reactor's core is composed of both high enriched and low enriched fuel and from high enriched to low enriched core conversion project will take place in year 2005. This paper presents the plans for drafting regulations on the safety analysis report updates for high enriched to low enriched core conversion of TR-2 reactor, the present regulatory structure of Turkey and licensing activities of nuclear facilities. (author)

Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

International Nuclear Information System (INIS)

Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

2016-01-01

Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1991-01-01

In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

Case study - Argentina

International Nuclear Information System (INIS)

Diaz, E.

1986-01-01

Antecedents and experience of nuclear activities in Argentina; the Atomic Energy Commission (CNEA). First development and research activities. Research reactors and radioisotopes plants. Health physics and safety regulations. - Feasibility studies for the first nuclear power plant. Awarding the first plant CNA I (Atucha I). Relevant data related to the different project stages. Plant performance. - Feasibility study for the second nuclear power plant. Awarding the second plant CNE (Central Nuclear Embalse). Relevant data related to established targets. Differences compared with the first station targets. Local participation. Plant performance. (orig./GL)

The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

International Nuclear Information System (INIS)

Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

2009-01-01

This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

Maximum credible accident analysis for TR-2 reactor conceptual design

International Nuclear Information System (INIS)

Manopulo, E.

1981-01-01

A new reactor, TR-2, of 5 MW, designed in cooperation with CEN/GRENOBLE is under construction in the open pool of TR-1 reactor of 1 MW set up by AMF atomics at the Cekmece Nuclear Research and Training Center. In this report the fission product inventory and doses released after the maximum credible accident have been studied. The diffusion of the gaseous fission products to the environment and the potential radiation risks to the population have been evaluated

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

Energy Technology Data Exchange (ETDEWEB)

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra; Su’ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Aji, Indarta K. [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia)

2016-03-11

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

The Chernobyl reactor accident. Pt. 1 and 2

International Nuclear Information System (INIS)

1986-06-01

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

Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

International Nuclear Information System (INIS)

2015-01-01

This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

2014-01-01

A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

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

International Nuclear Information System (INIS)

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

1995-06-01

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

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

International Nuclear Information System (INIS)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

Energy Technology Data Exchange (ETDEWEB)

Ramajo, Damian, E-mail: dramajo@santafe-conicet.gov.ar; Corzo, Santiago; Schiliuk, Nicolas; Nigro, Norberto

2013-12-15

A computational fluid dynamics (CFD) simulation of the reactor pressure vessel (RPV) of the pressurized heavy water reactor (PHWR) of 745 electrical MW Atucha II nuclear power plant was carried out. A three dimensional (3D) detailed model was employed to simulate coolant circuit considering the upper and lower plenums, the downcomer and the hot and cold legs. Control rods and coolant channel tubes at the upper plenum were included to quantify the mixing flow with more realism. The whole set of 451 coolant channels were modeled by means of a zero dimensional methodology. That is, the effect of each coolant channel was modeled through the introduction of a source point at the upper plenum and a sink point at the lower plenum. For each coupled sink/source points (SSP) the mass, momentum and energy balance were solved considering the local pressure difference and the temperature between the corresponding points where sinks and sources were placed. Based on this strategy, three models with increasingly level of approximation were implemented. For the first model the 451 coolant channels were reduced to only 57 pairs of SSP to represent all the coolant channels, concentrating the effect of several coolant channels in a unique pair of sink and source while taking into account geometric design details. For the second model, 225 pairs of SSP were introduced. Finally, for the third model each one of the 451 coolant channels were modeled by means of one pair of SSP. Depending on the coolant channel location, the radial power distribution and the pressure loss caused by the corresponding flow restrictor present by design were considered. Simulations carried out give insight in the complexity of the flow. As expected, the greater the details of the model the better the accuracy reached in the representation of the RPV behavior. In addition, the flow distributor located at the lower plenum showed to be very efficient since, the mass flow at each channel was found to be fairly

Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

2013-01-01

This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. 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 core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

2010-12-15

The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

Recent advances in the utilization and the irradiation technology of the refurbished BR2 reactor

International Nuclear Information System (INIS)

Dekeyser, J.; Benoit, P.; Decloedt, C.; Pouleur, Y.; Verwimp, A.; Weber, M.; Vankeerberghen, M.; Ponsard, B.

1999-01-01

Operation and utilization of the materials testing reactor BR2 at the Belgian Nuclear Research Centre (SCK·CEN) has since its start in 1963 always followed closely the needs and developments of nuclear technology. In particular, a multitude of irradiation experiments have been carried out for most types of nuclear power reactors, existing or under design. Since the early 1990s and increased focus was directed towards more specific irradiation testing needs for light water reactor fuels and materials, although other areas of utilization continued as well (e.g. fusion reactor materials, safety research, ...), including also the growing activities of radioisotope production and silicon doping. An important milestone was the decision in 1994 to implement a comprehensive refurbishment programme for the BR2 reactor and plant installations. The scope of this programme comprised very substantial studies and hardware interventions, which have been completed in early 1997 within planning and budget. Directly connected to this strategic decision for reactor refurbishment was the reinforcement of our efforts to requalify and upgrade the existing irradiation facilities and to develop advanced devices in BR2 to support emerging programs in the following fields: - LWR pressure vessel steel, - LWR irradiation assisted stress corrosion cracking (IASCC), - reliability and safety of high-burnup LWR fuel, - fusion reactor materials and blanket components, - fast neutron reactor fuels and actinide burning, - extension and diversification of radioisotope production. The paper highlights these advances in the areas of BR2 utilisation and the ongoing development activities for the required new generation of irradiations devices. (author)

An improved thermal-hydraulic modeling of the Jules Horowitz Reactor using the CATHARE2 system code

Energy Technology Data Exchange (ETDEWEB)

Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

2017-01-15

Highlights: • An improved thermal-hydraulic modeling of the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during loss of flow accident. • The heat exchanger approach gives more realistic and less conservative results. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support current and future nuclear reactor designs. The reactor is under construction at the CEA Cadarache research center in southern France and is expected to achieve first criticality at the end of this decade. This paper presents an improved thermal-hydraulic modeling of the reactor using solely CATHARE2 system code. Up to now, the CATHARE2 code was simulating the full reactor with a simplified approach for the core and the boundary conditions were transferred into the three-dimensional FLICA4 core simulation. A new more realistic methodology is utilized to analyze the thermal-hydraulic simulation of the reactor during a loss of flow accident.

Hybrid reactors

International Nuclear Information System (INIS)

Moir, R.W.

1980-01-01

The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

A level III PSA for the inherently safe CAREM-25 nuclear power station

International Nuclear Information System (INIS)

Baron, Jorge H.; Nunez McLeod, J.; Rivera, S.S.

2000-01-01

A Level III PSA has been performed for the inherently safe CAREM-25 nuclear power station, as a requirement for licensing according to argentinian regulations. The CAREM-25 project is still at a detailed design state, therefore only internal events have been considered, and a representative site has been assumed for dose estimations. Several conservative hypothesis have been formulated, but even so an overall core melt frequency of 2.3E -5 per reactor year has been obtained. The risk estimations comply with the regulations. The risk values obtained are compared to the 700MW(e) nuclear power plant Atucha II PSA result, showing an effective risk reduction not only in the severe accident probability but alto in the consequence component of the risk estimation. (author)

Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966; 2. Jugoslovenski simpozijum iz reaktorske fizike, Deo 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

Energy Technology Data Exchange (ETDEWEB)

NONE

1966-07-01

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.

Problems of nuclear reactor safety. Vol. 2

International Nuclear Information System (INIS)

Goncharov, L.A.

1995-01-01

Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

International Nuclear Information System (INIS)

Aziz, M.; El Messiry, A.M.

2000-01-01

The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

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

International Nuclear Information System (INIS)

1978-05-01

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

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

International Nuclear Information System (INIS)

Sinha, V.P.; Hegde, P.V.; Prasad, G.J.; Pal, S.; Mishra, G.P.

2012-01-01

CERMET fuel with either PuO 2 or enriched UO 2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR’s). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R and D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO 2 dispersed in uranium metal matrix pellets for three different compositions i.e. U–20 wt%UO 2 , U–25 wt%UO 2 and U–30 wt%UO 2 . It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U–UO 2 compositions.

Development of UO2/PuO2 dispersed in uranium matrix CERMET fuel system for fast reactors

Science.gov (United States)

Sinha, V. P.; Hegde, P. V.; Prasad, G. J.; Pal, S.; Mishra, G. P.

2012-08-01

CERMET fuel with either PuO2 or enriched UO2 dispersed in uranium metal matrix has a strong potential of becoming a fuel for the liquid metal cooled fast breeder reactors (LMR's). In fact it may act as a bridge between the advantages and disadvantages associated with the two extremes of fuel systems (i.e. ceramic fuel and metallic fuel) for fast reactors. At Bhabha Atomic Research Centre (BARC), R & D efforts are on to develop this CERMET fuel by powder metallurgy route. This paper describes the development of flow sheet for preparation of UO2 dispersed in uranium metal matrix pellets for three different compositions i.e. U-20 wt%UO2, U-25 wt%UO2 and U-30 wt%UO2. It was found that the sintered pellets were having excellent integrity and their linear mass was higher than that of carbide fuel pellets used in Fast Breeder Test Reactor programme (FBTR) in India. The pellets were characterized by X-ray diffraction (XRD) technique for phase analysis and lattice parameter determination. The optical microstructures were developed and reported for all the three different U-UO2 compositions.

Utilization of the SLOWPOKE-2 research reactor

International Nuclear Information System (INIS)

Lalor, G.C.

2001-01-01

SLOWPOKEs are typically low power research reactors that have a limited number of applications. However, a significant range of NAA can be performed with such reactors. This paper describes a SLOWPOKE-based NAA program that is performing a valuable series of studies in Jamaica, including geological mapping and pollution assessment. (author)

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

International Nuclear Information System (INIS)

1982-01-01

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

Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

International Nuclear Information System (INIS)

Markovic, H.

1962-07-01

The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

International Nuclear Information System (INIS)

Souza Mendes, J.E. de

1987-01-01

Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

Application of the neutron noise analysis technique in nuclear power plants

International Nuclear Information System (INIS)

Lescano, Victor H.; Wentzeis, Luis M.

1999-01-01

Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

International Nuclear Information System (INIS)

Barelli, L.; Bidini, G.; Gallorini, F.

2016-01-01

Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

Directory of Open Access Journals (Sweden)

Abdul Wahid

2015-06-01

Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

Operating reactors licensing actions summary. Volume 5, No. 2

International Nuclear Information System (INIS)

1985-04-01

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

Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

Energy Technology Data Exchange (ETDEWEB)

Nilsson, R; Aalto, E

1964-09-15

Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

Microflow photochemistry: UVC-induced [2 + 2]-photoadditions to furanone in a microcapillary reactor

Directory of Open Access Journals (Sweden)

Sylvestre Bachollet

2013-10-01

Full Text Available [2 + 2]-Cycloadditions of cyclopentene and 2,3-dimethylbut-2-ene to furanone were investigated under continuous-flow conditions. Irradiations were conducted in a FEP-microcapillary module which was placed in a Rayonet chamber photoreactor equipped with low wattage UVC-lamps. Conversion rates and isolated yields were compared to analogue batch reactions in a quartz test tube. In all cases examined, the microcapillary reactor furnished faster conversions and improved product qualities.

A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

International Nuclear Information System (INIS)

Abd El-Rahman, M.F.

2004-01-01

Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

Thermal design of heat-exchangeable reactors using a dry-sorbent CO2 capture multi-step process

International Nuclear Information System (INIS)

Moon, Hokyu; Yoo, Hoanju; Seo, Hwimin; Park, Yong-Ki; Cho, Hyung Hee

2015-01-01

The present study proposes a multi-stage CO 2 capture process that incorporates heat-exchangeable fluidized-bed reactors. For continuous multi-stage heat exchange, three dry regenerable sorbents: K 2 CO 3 , MgO, and CaO, were used to create a three-stage temperature-dependent reaction chain for CO 2 capture, corresponding to low (50–150 °C), middle (350–650 °C), and high (750–900 °C) temperature stages, respectively. Heat from carbonation in the high and middle temperature stages was used for regeneration for the middle and low temperature stages. The feasibility of this process is depending on the heat-transfer performance of the heat-exchangeable fluidized bed reactors as the focus of this study. The three-stage CO 2 capture process for a 60 Nm 3 /h CO 2 flow rate required a reactor area of 0.129 and 0.130 m 2 for heat exchange between the mid-temperature carbonation and low-temperature regeneration stages and between the high-temperature carbonation and mid-temperature regeneration stages, respectively. The reactor diameter was selected to provide dense fluidization conditions for each bed with respect to the desired flow rate. The flow characteristics and energy balance of the reactors were confirmed using computational fluid dynamics and thermodynamic analysis, respectively. - Highlights: • CO 2 capture process is proposed using a multi-stage process. • Reactor design is conducted considering heat exchangeable scheme. • Reactor surface is designed by heat transfer characteristics of fluidized bed

Increased SRP reactor power

International Nuclear Information System (INIS)

MacAfee, I.M.

1983-01-01

Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

Core design calculations of IRIS reactor using modified CORD-2 code package

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

2002-01-01

Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

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

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

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

BR2 reactor: medical and industrial applications

International Nuclear Information System (INIS)

Ponsard, B.

2005-01-01

The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

Possible future roles for fast breeder reactors Part 1 and 2

International Nuclear Information System (INIS)

1978-06-01

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

A regression approach for Zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

International Nuclear Information System (INIS)

Elliot, A.J.; Stuart, C.R.

2008-06-01

An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

Design and computational analysis of passive siphon breaker for 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Yue Zhiting; Song Yunpeng; Liu Xingmin; Zou Yao; Wu Yuanyuan

2014-01-01

Based on safety considerations, a passive siphon breaker will be added to the primary cooling system of 49-2 Swimming Pool Reactor (SPR). With the breaker location determined, the capability of siphon breakers with diameters of 1.5 cm and 2.0 cm was calculated and analyzed respectively by RELAP5/MOD3.3 code. The results show that in the condition of large break loss of coolant accident these two sizes of siphon breakers are able to break the siphon phenomena, and maintain the pool water level above the reactor core when the reactor and the pump are shutdown. In the end, to be conservative, the siphon breaker with diameter of 2.0 cm is adopted. (authors)

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.

Direct In Situ Quantification of HO2 from a Flow Reactor.

Science.gov (United States)

Brumfield, Brian; Sun, Wenting; Ju, Yiguang; Wysocki, Gerard

2013-03-21

The first direct in situ measurements of hydroperoxyl radical (HO2) at atmospheric pressure from the exit of a laminar flow reactor have been carried out using mid-infrared Faraday rotation spectroscopy. HO2 was generated by oxidation of dimethyl ether, a potential renewable biofuel with a simple molecular structure but rich low-temperature oxidation chemistry. On the basis of the results of nonlinear fitting of the experimental data to a theoretical spectroscopic model, the technique offers an estimated sensitivity of reactor exit temperature range of 398-673 K. Accurate in situ measurement of this species will aid in quantitative modeling of low-temperature and high-pressure combustion kinetics.

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Status and perspective of development of cold moderators at the IBR-2 reactor

International Nuclear Information System (INIS)

Kulikov, S; Shabalin, E

2012-01-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Status and perspective of development of cold moderators at the IBR-2 reactor

Science.gov (United States)

Kulikov, S.; Shabalin, E.

2012-03-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

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

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

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.)

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

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

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

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

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

International Nuclear Information System (INIS)

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

1990-01-01

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

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

Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

International Nuclear Information System (INIS)

Skogen, F.B.; Stout, R.B.

1977-01-01

Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

Characterization of fuel distribution in the Three Mile Island Unit 2 (TMI-2) reactor system by neutron and gamma-ray dosimetry

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McNeece, J.P.; Kaiser, B.J.; McElroy, W.N.

1984-01-01

Neutron and gamma-ray dosimetry are being used for nondestructive assessment of the fuel distribution throughout the Three Mile Island Unit 2 (TMI-2) reactor core region and primary cooling system. The fuel content of TMI-2 makeup and purification Demineralizer A has been quantified with Si(Li) continuous gamma-ray spectrometry and solid-state track recorder (SSTR) neutron dosimetry. For fuel distribution characterization in the core region, results from SSTR neutron dosimetry exposures in the TMI-2 reactor cavity are presented. These SSTR results are consistent with the presence of a significant amount of fuel debris, equivalent to several fuel assemblies or more, lying at the bottom of the reactor vessel. (Auth.)

Estimation of power feedback parameters of the IBR-2M reactor by square wave reactivity

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Popov, A.K.; Sumkhuu, D.

2016-01-01

Parameters of the IBR-2M reactor power feedback (PFB) are estimated based on the analysis of power transients caused by deliberate square wave reactivity when the pulsed reactor operates in the self-regulation mode. The PFB of the IBR-2M is described by three linear first-order differential equations. Two components of the PFB are responsible for the negative feedback and one, for the positive. The overall feedback is negative, i.e., it has a stabilizing effect for the operation of the reactor. The slowest negative component of the PFB is probably caused by heating of the fuel. Periodically repeated in the process of exploitation, estimation of the PFB parameters is one of the methods to ensure safety operation of the reactor. [ru

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

TOKMINA, Toroidal Magnetic Field Minimization for Tokamak Fusion Reactor. TOKMINA-2, Total Power for Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Hatch, A.J.

1975-01-01

1 - Description of problem or function: TOKMINA finds the minimum magnetic field, Bm, required at the toroidal coil of a Tokamak type fusion reactor when the input is beta(ratio of plasma pressure to magnetic pressure), q(Kruskal-Shafranov plasma stability factor), and y(ratio of plasma radius to vacuum wall radius: rp/rw) and arrays of PT (total thermal power from both d-t and tritium breeding reactions), Pw (wall loading or power flux) and TB (thickness of blanket), following the method of Golovin, et al. TOKMINA2 finds the total power, PT, of such a fusion reactor, given a specified magnetic field, Bm, at the toroidal coil. 2 - Method of solution: TOKMINA: the aspect ratio(a) is minimized, giving a minimum value for Bm. TOKMINA2: a search is made for PT; the value of PT which minimizes Bm to the required value within 50 Gauss is chosen. 3 - Restrictions on the complexity of the problem: Input arrays presently are dimensioned at 20. This restriction can be overcome by changing a dimension card

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

International Nuclear Information System (INIS)

Ellis, R.J.

2000-01-01

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

Report on the operation in 1973 of the FR 2 research reactor

International Nuclear Information System (INIS)

Moeller, I.; Steiger, W.

1975-04-01

Also in 1973, the heavy-water moderated research and testing reactor FR 2 was operated to schedule at 44 MW nominal power. Again, the availability of the plant was slightly improved. Experimental utilization through instrumented irradiation capsules strongly increased as compared to the previous year. Up to 16 capsule test rigs at a time were inserted in the reactor. As to the beam tube experiments, up to 13 experiments covering a total of 18 test rigs were conducted simultaneously at the 12 reasonably usable beam holes. At the beginning of the year all of the positions available were occupied by 5 loop experiments. Isotope production reached its highest value with a total of 2,372 irradiated capsules (1.3% more than the year before). Some remarkable figures characterized the year 1973: On August 16, 1973 ten years of full power operation at a nominal power of 12 and 44 MW, respectively, had been reached. On July 24, 1973 the 50,000th isotope irradiation was performed in the reactor and on December 26, 1973 a total energy release of 100,000 MWd was recorded. Moreover, the 125,000th visitor of the reactor was welcomed on December 6, 1973. (orig./UA) [de

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

International Nuclear Information System (INIS)

Doerge, D.H.; Miller, R.L.; Scotti, K.S.

1986-05-01

This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

Energy Technology Data Exchange (ETDEWEB)

Dougherty, D.; Adams, J. W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

International Nuclear Information System (INIS)

Dougherty, D.; Adams, J.W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor

International Nuclear Information System (INIS)

Finch, D.R.

1965-01-01

1 - Description of problem or function: HERESY3 solves the two- dimensional, few-group, static reactor eigenvalue problem using the heterogeneous (source-sink or Feinburg-Galanin) formalism. The solution yields the reactor k-effective and absorption reaction rates for each rod normalized to the most absorptive rod in the thermal level. Epithermal fissions are allowed at each resonance level, and lattice-averaged values of thermal utilization, resonance escape probability, thermal and resonance eta values, and the fast fission factor are calculated. Kernels in the calculation are based on age-diffusion theory. Both finite reactor lattices and infinitely repeating reactor super-cells may be calculated. Rod parameters may be calculated by several internal options, and a direct interface is provided to a HAMMER system (NESC Abstract 277) lattice library tape to obtain cell parameters. Criticality searches are provided on thermal utilization, thermal eta, and axial leakage buckling. 2 - Method of solution: Direct power iteration on matrix form of the heterogeneous critical equation is used. 3 - Restrictions on the complexity of the problem: Maxima of - 50 flux/geometry symmetry positions; 20 physically different assemblies; 9 resonance levels; 5000 rod coordinate positions

Exploitation of BEPU Approach for the Licensing Process

International Nuclear Information System (INIS)

D'Auria, F.; Petruzzi, A.; Muellner, A.N.; Mazzantini, O.

2011-01-01

Within the licensing process of the Atucha II PHWR (Pressurized Heavy Water Reactor) the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: a) the selection of PIE (Postulated Initiating Events) and, b) the analysis by best estimate models supported by uncertainty evaluation. The key elements of the approach are: 1) availability of qualified computational tools including suitable uncertainty method; 2) demonstration of quality; 3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified. (authors)

G2 and G3 reactors design; Description des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

Investigation of hydrogen-burn damage in the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Alvares, N.J.; Beason, D.G.; Eidem, G.R.

1982-06-01

About 10 hours after the March 28, 1979 Loss-of-Coolant Accident began at Three Mile Island Unit 2, a hydrogen deflagration of undetermined extent occurred inside the reactor building. Examinations of photographic evidence, available from the first fifteen entries into the reactor building, yielded preliminary data on the possible extent and range of hydrogen burn damage. These data, although sparse, contributed to development of a possible damage path and to an estimate of the extent of damage to susceptible reactor building items. Further information gathered from analysis of additional photographs and samples can provide the means for estimating hydrogen source and production rate data crucial to developing a complete understanding of the TMI-2 hydrogen deflagration. 34 figures

A review of the probabilistic safety assessment application to the TR-2 research reactor

International Nuclear Information System (INIS)

Goektepe, G.; Adalioglu, U.; Anac, H.; Sevdik, B.; Menteseoglu, S.

2001-01-01

A review of the Probabilistic Safety Assessment (PSA) to the TR-2 Research Reactor is presented. The level 1 PSA application involved: selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development, dependent failure analysis. Each of the steps of the analysis given above is reviewed briefly with highlights from the selected results. PSA application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. Insights gained from the application of PSA methodology to the TR-2 research reactor led to a significant safety review of the system

Solid-state track recorder neutron dosimetry in the Three-Mile Island Unit-2 reactor cavity

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.

1985-04-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that there are at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1986-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity, for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. This existence of significant neutron streaming also explains the high count rate observed with the source range monitors that are located in the TMI-2 reactor cavity. (author)

Degradation of gas-phase trichloroethylene over thin-film TiO2 photocatalyst in multi-modules reactor

International Nuclear Information System (INIS)

Kim, Sang Bum; Lee, Jun Yub; Kim, Gyung Soo; Hong, Sung Chang

2009-01-01

The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO 2 . A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

A gas-phase reactor powered by solar energy and ethanol for H2 production

International Nuclear Information System (INIS)

Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

2014-01-01

In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Delayed hydride cracking of Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

2007-01-01

Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

The text of the agreement of 15 July 1981 between Argentina and the Agency for the application of Safeguards in connection with contracts concluded between the Comision Nacional de Energia Atomica Argentina and the Kraftwerk Union AG (Federal Republic of Germany) for the supply of the Atucha II Nuclear Power Plant

International Nuclear Information System (INIS)

1995-01-01

The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Agreement of 15 July 1981 between Argentina and the IAEA for the application of safeguards in connection with contracts concluded between the Comision Nacional de Energia Atomic Argentina and the Kraftwerk Union AG (Federal Republic of Germany) for the supply of the Atucha II Nuclear Power Plant has been suspended

Neutron dosimetry in the Three-Mile Island Unit 2 reactor cavity with solid-state track recorders

International Nuclear Information System (INIS)

Gold, R.; Roberts, J.H.; Ruddy, F.H.; Preston, C.C.; McElroy, W.N.; Rao, S.V.; Greenborg, J.; Fricke, V.R.

1985-01-01

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI-2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity

A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

Directory of Open Access Journals (Sweden)

Hwanyeal Yu

2015-12-01

Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

Analysis of key hardware factors and countermeasure for restricting 49-2 swimming pool reactor lifetime

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Yang Xiao; Wang Yiwei; Wang Zhanwen

2013-01-01

Safe operation is the most important factor to determine the lifetime of aged 49-2 swimming pool reactor. In this paper, the hardware factors of lifetime were analyzed, such as the pool concrete aging, corrosion of aluminum container and primary coolant system, and graphite swelling etc., and then the corresponding measures such as surveillance, prevention and maintenance were purposed. The results show that 49-2 swimming pool reactor can continue to operate safely due to that container is safe under 8 degree earthquake, the reactor is safe on flood level of once per millennium, adding dam break, and the ageing condition of primary coolant system and container is acceptable. (authors)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

additional consideration should be required in nuclear design and fuel treating facilities due to reactivity coefficient being shifted to the plus side, larger neutron yield and increased heat source caused by MA loading. (2) Confirmation of TRU burning reactor core concepts. The core specification of sodium cooled-nitride fueled TRU burning large reactor was designed based on commercial type fast reactor (sodium cooled nitride fueled large fast reactor, 38000 MWt) which was designed in the feasibility studies on commercialized fast reactor cycle system. The composition of MAs from LWR's spent fuel was supposed. MA content in the core fuel is settled to 60 wt% based on the JAERI's design in order to maximize the MA transmutation amount. We need to exchange 25% of core fuel with zirconium hydride (ZrH 1.6 ) to attain Doppler coefficient being equivalent to that of the conventional type commercial fast reactor loaded 5 wt% MA. Furthermore, this reactor could transmute MAs produced in forty-eight sodium cooled nitride fueled large fast reactors generating the same output. In order to investigate the dependency of MA transmutation characteristics on the reactor output, 1200 MWt TRU burning middle or small reactor core concept was designed. This core was settled by reducing the number of core fuel assemblies from that of TRU burning large reactor designed above. MA transmutation rate in this core is smaller than that in the TRU burning large reactor core because the neutron flux of this core becomes smaller than that of the TRU burning large reactor core due to the higher Pu enrichment. (3) Comparison between TRU burning reactor and conventional type commercial fast reactor. MA transmutation and nuclear characteristics of the sodium cooled nitride fuel commercial type fast reactor loaded 5 wt%MA were evaluated and compared with those of TRU burning large reactor designed in (2). The commercial type fast reactor could only transmute MAs produced in seven sodium cooled nitride

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

International Nuclear Information System (INIS)

Collins, N.; Lafreniere, P.

1994-01-01

In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

Cutting and dismantling of the South West ladder of the Atucha I nuclear power plant

International Nuclear Information System (INIS)

Anasco, Roberto

2006-01-01

The metallic ladder built in stainless steel was used originally to check the welding of the reactor pressure vessel. It was located between the thermal insulation and reactor pressure vessel. Because of a failure in the mechanism, which let the ladder runs around the vessel, it had to be removed. A special tool remotely operated was designed to make different cuts in the bottom of the structure in a very high radioactive location [es

Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

International Nuclear Information System (INIS)

Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik

2015-01-01

A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized

VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions

International Nuclear Information System (INIS)

Jackson, J.F.; Nicholson, R.B.; Weber, D.P.

1980-01-01

1 - Description of problem or function: VENUS-2 is an improved edition of the VENUS fast-reactor disassembly program. It is a two- dimensional (r-z) coupled neutronics-hydrodynamics code that calculates the dynamic behavior of an LMFBR during a prompt-critical disassembly excursion. It calculates the power history and fission energy release as well as the space-time histories of the fuel temperatures, core material pressures, and core material motions. Reactivity feedback effects due to Doppler broadening and reactor material motion are taken into account. 2 - Method of solution: The power and energy release are calculated using a point-kinetics formulation with up to six delayed neutron groups. The reactivity is a combination of an input driving function and feedback effects due to Doppler broadening and material motion. An adiabatic model is used to calculate the temperature increase throughout the reactor based on an initial temperature distribution and power profile provided as input data. These temperatures are, in turn, converted to fuel pressures through one of several equation of state options provided. The material motion that results from the pressure buildup is calculated by a direct finite difference solution of a set of two-dimensional (r-z) hydrodynamics equations. This is done in Lagrangian coordinates. The reactivity change associated with this motion is calculated by first-order perturbation theory. The displacements are also used to adjust the fuel densities as required for the density dependent equation-of- state option. An automatic time-step-size selection scheme is provided. 3 - Restrictions on the complexity of the problem: VENUS-2 is written so that the dimensions of the storage arrays can be readily changed to accommodate a broad range of problem sizes. In the base version, the total number of mesh intervals is restricted such that (NR+3)*(NZ+3) is less than 700, where NR and NZ are the total number of mesh intervals in the r and z

Energy Multiplier Module (EM{sup 2}) - advanced small modular reactor for electricity generation

Energy Technology Data Exchange (ETDEWEB)

Bertch, T.; Schleicher, R.; Choi, H.; Rawls, J., E-mail: timothy.bertch@ga.com [General Atomics, San Diego, California (United States)

2013-07-01

In order to provide cost effective nuclear energy in other than large reactor, large grid applications, fission technology needs to make further advances. 'Convert and burn' fast reactors offer long life cores, improved fuel utilization, reduced waste and other benefits while achieving cost effective energy production in a smaller reactor. General Atomics' Energy Multiplier Module (EM{sup 2}), a helium-cooled compact fast reactor that augments its fissile fuel load with either depleted uranium (DU) or used nuclear fuel (UNF). The convert and burn in-situ provides 250 MWe with a 30 year core life. High temperature provides a simple, high efficiency direct cycle gas turbine which along with modular construction, fewer systems, road shipment and minimum on site construction support cost effectiveness. Additional advantages in fuel cycle, non-proliferation and siting flexibility and its ability to meet all safety requirements make for an attractive power source, especially in remote and small grid regions. (author)

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

International Nuclear Information System (INIS)

Perez, S.S.

1987-01-01

The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

Reactor noise analysis of experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Ohtani, Hideji; Yamamoto, Hisashi

1980-01-01

As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

Application of 2DOF controller for reactor power control. Verification by numerical simulation

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Suzuki, Katsuo

1996-09-01

In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)

Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2007-01-01

The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

International Nuclear Information System (INIS)

1981-06-01

This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

Degradation of gas-phase trichloroethylene over thin-film TiO{sub 2} photocatalyst in multi-modules reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Bum [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Lee, Jun Yub, E-mail: ljy02191@hanafos.com [Power Engineering Research Institute, Korea Power Engineering Company, Inc. (Korea, Republic of); Kim, Gyung Soo [New and Renewable Energy Team, Environment and Energy Division, Korea Institute of Industrial Technology (Korea, Republic of); Hong, Sung Chang [Department of Environmental Engineering, Kyonggi University (Korea, Republic of)

2009-07-30

The present paper examined the photocatalytic degradation (PCD) of gas-phase trichloroethylene (TCE) over thin-film TiO{sub 2}. A large-scale treatment of TCE was carried out using scale-up continuous flow photo-reactor in which nine reactors were arranged in parallel and series. The parallel or serial arrangement is a significant factor to determine the special arrangement of whole reactor module as well as to compact the multi-modules in a continuous flow reactor. The conversion of TCE according to the space time was nearly same for parallel and serial connection of the reactors.

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

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

CERN Document Server

Chiba, G

2002-01-01

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

Reactor Physics Training

International Nuclear Information System (INIS)

Baeten, P.

2007-01-01

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

Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

International Nuclear Information System (INIS)

Potapov, I.A.; Serebrov, A.P.

2001-01-01

The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH 2 ) and liquid deuterium (LD 2 ) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

International Nuclear Information System (INIS)

1978-07-01

The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

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

Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

Energy Technology Data Exchange (ETDEWEB)

Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

2014-09-15

Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

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

Mass transfer of ammonia escape and CO2 absorption in CO2 capture using ammonia solution in bubbling reactor

International Nuclear Information System (INIS)

Ma, Shuangchen; Chen, Gongda; Zhu, Sijie; Han, Tingting; Yu, Weijing

2016-01-01

Highlights: • Mass transfer coefficient models of ammonia escape were built. • Influences of temperature, inlet CO 2 and ammonia concentration were studied. • Mass transfer coefficients of ammonia escape and CO 2 absorption were obtained. • Studies can provide the basic data as a reference guideline for process application. - Abstract: The mass transfer of CO 2 capture using ammonia solution in the bubbling reactor was studied; according to double film theory, the mass transfer coefficient models and interface area model were built. Through our experiments, the overall volumetric mass transfer coefficients were obtained, while the interface areas in unit volume were estimated. The volumetric mass transfer coefficients of ammonia escaping during the experiment were 1.39 × 10 −5 –4.34 × 10 −5 mol/(m 3 s Pa), and the volumetric mass transfer coefficients of CO 2 absorption were 2.86 × 10 −5 –17.9 × 10 −5 mol/(m 3 s Pa). The estimated interface area of unit volume in the bubbling reactor ranged from 75.19 to 256.41 m 2 /m 3 , making the bubbling reactor a viable choice to obtain higher mass transfer performance than the packed tower or spraying tower.

SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

Science.gov (United States)

Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

2016-04-01

Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

A novel condensation reactor for efficient CO2 to methanol conversion for storage of renewable electric energy

NARCIS (Netherlands)

Bos, Martin Johan; Brilman, Derk Willem Frederik

2015-01-01

A novel reactor design for the conversion of CO2 and H2 to methanol is developed. The conversion limitations because of thermodynamic equilibrium are bypassed via in situ condensation of a water/methanol mixture. Two temperatures zones inside the reactor ensure optimal catalyst activity (high

Quality assurance in the project of RECH-2 research reactor

International Nuclear Information System (INIS)

Goycolea Donoso, C.; Nino de Zepeda Schele, A.

1989-01-01

The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

Set of rules SOR 2 licensing of nuclear reactors

International Nuclear Information System (INIS)

1976-05-01

This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

Failure Analysis

International Nuclear Information System (INIS)

Iorio, A.F.; Crespi, J.C.

1987-01-01

After ten years of operation at the Atucha I Nuclear Power Station a gear belonging to a pressurized heavy water reactor refuelling machine, failed. The gear box was used to operate the inlet-outlet heavy-water valve of the machine. Visual examination of the gear device showed an absence of lubricant and that several gear teeth were broken at the root. Motion was transmitted with a speed-reducing device with controlled adjustable times in order to produce a proper fitness of the valve closure. The aim of this paper is to discuss the results of the gear failure analysis in order to recommend the proper solution to prevent further failures. (Author)

Measurement of thermal conductivity of sintered UO{sub 2} in the reactor; Merenje toplotne provodljivosti sinterovanog UO{sub 2} u reaktoru

Energy Technology Data Exchange (ETDEWEB)

Katanic, J; Stevanovic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1965-10-15

Thermal conductivity is considered one of the fundamental properties of sintered UO{sub 2} fuel. Samples should be tested under real core conditions. This paper covers the methods and instruments for thermal conductivity measurement of UO{sub 2} samples in the reactor core, measurements outside the core under conditions similar to those in the core and outside the core after irradiation. Fuel samples are placed in capsules for irradiation in the reactor in-core loops.

Geomechanical Analysis of Underground Coal Gasification Reactor Cool Down for Subsequent CO2 Storage

Science.gov (United States)

Sarhosis, Vasilis; Yang, Dongmin; Kempka, Thomas; Sheng, Yong

2013-04-01

Underground coal gasification (UCG) is an efficient method for the conversion of conventionally unmineable coal resources into energy and feedstock. If the UCG process is combined with the subsequent storage of process CO2 in the former UCG reactors, a near-zero carbon emission energy source can be realised. This study aims to present the development of a computational model to simulate the cooling process of UCG reactors in abandonment to decrease the initial high temperature of more than 400 °C to a level where extensive CO2 volume expansion due to temperature changes can be significantly reduced during the time of CO2 injection. Furthermore, we predict the cool down temperature conditions with and without water flushing. A state of the art coupled thermal-mechanical model was developed using the finite element software ABAQUS to predict the cavity growth and the resulting surface subsidence. In addition, the multi-physics computational software COMSOL was employed to simulate the cavity cool down process which is of uttermost relevance for CO2 storage in the former UCG reactors. For that purpose, we simulated fluid flow, thermal conduction as well as thermal convection processes between fluid (water and CO2) and solid represented by coal and surrounding rocks. Material properties for rocks and coal were obtained from extant literature sources and geomechanical testings which were carried out on samples derived from a prospective demonstration site in Bulgaria. The analysis of results showed that the numerical models developed allowed for the determination of the UCG reactor growth, roof spalling, surface subsidence and heat propagation during the UCG process and the subsequent CO2 storage. It is anticipated that the results of this study can support optimisation of the preparation procedure for CO2 storage in former UCG reactors. The proposed scheme was discussed so far, but not validated by a coupled numerical analysis and if proved to be applicable it could

Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

International Nuclear Information System (INIS)

Wu Yuanyuan; Liu Tiancai; Sun Wei

2012-01-01

The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

Status of IVO-FR2-Vg7 experiment for irradiation of fast reactor fuel rods

International Nuclear Information System (INIS)

Elbel, H.; Kummerer, K.; Bojarsky, K.; Lopez Jimenez, J.; Otero de la Gandara, J.L.

1979-01-01

Report on the Seminar celebrated in Madrid between KfK (Karlsruhe) and JEN (Madrid) concerning a Joint Irradiation Program of Fast Reactor Fuel Rods. The design of fuel rods in general is defined, and, in particular of those with a density 94% DT and diameter 7.6 mm up to a burn-up of 7% FIMA, to be irradiated in the FR2 Reactor (Karlsruhe). Together with the design of NaK and single-wall capsules used in this irradiation, other possibilities of irradiation in the reactor will also be described. (auth.)

CANDU reactors with reactor grade plutonium/thorium carbide fuel

Energy Technology Data Exchange (ETDEWEB)

Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

2011-08-15

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

Revision of fast reactor group constant set JFS-3-J2

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

UCLA program in reactor studies: The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

N2O Catalytic Decomposition – from Laboratory Experiment to Industry Reactor

Czech Academy of Sciences Publication Activity Database

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

2012-01-01

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

Nuclear Reactor RA Safety Report, Vol. 4, Reactor

International Nuclear Information System (INIS)

1986-11-01

RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

Alteration of installation of reactors (alteration of No.1 and No.2 reactor facilities) in Oi Power Station, Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

1984-01-01

The Nuclear Safety Commission reported to the Minister of International Trade and Industry on October 27, 1983, that the technical capability was recognized to be adequate, and the safety after the alteration of the installation of reactors was judged to be ensured. At the time of deliberation, the guidelines for examining the safety design and safety evaluation of LWR facilities for power generation were used. Regarding the change of the degree of enrichment of replacement fuel from 3.2 to 3.4 wt.%, the limiting conditions are satisfied in the replacement core, and the nuclear design is appropriate. Eight test fuel assemblies using UO 2 pellets containing gadolinia are charged in the core of No.2 reactor, and the irradiation of two cycles is carried out. As the result of the safety examination regarding this test, the propriety of the nuclear design and mechanical design of the test fuel assemblies was confirmed. This alteration does not exert influence on the result of safety analysis made so far. This report was decided by the Committee on Examination of Reactor Safety based on the conclusion of No.26 subcommittee. (Kako, I.)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

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.

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)