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Sample records for afrri-triga reactor facility

  1. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    International Nuclear Information System (INIS)

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

  2. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  3. Analysis of cocked fuel elements in the AFRRI TRIGA Mark-F reactor

    International Nuclear Information System (INIS)

    The Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F pulsing reactor has experienced eight cocked fuel elements during the period 5 November 1974 through 17 February 1982. Although there are no adverse health and safety consequences associated with their occurrence and there is no credible potential for system damage, cocked TRIGA fuel elements do cause inconvenience to the reactor staff and a temporary delay in operations. This paper presents the history of cocked TRIGA fuel elements at AFRRI, discusses possible mechanisms for their occurrence, and outlines a plan to isolate and ultimately determine their actual cause

  4. Calculations of radiation fields and monkey mid-head and mid-thorax responses in AFRRI-TRIGA reactor facility experiments

    International Nuclear Information System (INIS)

    A computational study was performed to characterize the radiation exposure fields and the mid-head and mid-thorax response functions for monkeys irradiated in the Armed Forces Radiobiological Research Institute (AFRRI) reactor exposure facilities. Discrete ordinates radiation transport calculations were performed in one-dimensional spherical geometry to obtain the energy spectra of the neutrons and gamma rays entering the room through various spectrum modifiers and reaching the irradiation position. Adjoint calculations performed in two-dimensional cylindrical geometry yielded the mid-head and mid-thorax response functions, which were then folded with flux spectra to obtain the monkey mid-head and mid-thorax doses (kerma rates) received at the irradiation position. The results of the study are presented both as graphs and as tables. The resulting spectral shapes compared favorably with previous work; however, the magnitudes of the fluxes did not. The differences in the magnitudes may be due to the normalization factor used

  5. The AFRRI TRIGA reactor: a summary of applications in mouse studies - 345

    International Nuclear Information System (INIS)

    The AFRRI TRIGA reactor was used to simulate nuclear weapon mixed-field radiation injuries with and without additional tissue trauma. The severity of reactor-produced mixed-field radiations over that of γ-photon irradiation was evaluated in mice. Lethal doses (LDs) to 50% of groups of mice were determined for marrow cell (LD50/30, the dose required to kill 50% of the subjects within 30 days) and intestinal cell (LD50/6, the dose required to kill 50% of the subjects within 6 days) injury. As neutron (n) proportions in the total (t) radiation dose (Dn/Dt) increased LD values decreased. Relative biological effectiveness (RBE) values for reactor-generated Dn/Dt used 60Co γ photons and 250-kVp x-rays as reference standards. RBEs for irradiated mice increased as Dn/Dt increased and was further increased by wound trauma. Compared to γ-photon irradiation, mixed-field irradiation delayed wound closure times 25% to 50%. WR-151327 (200 mg/kg), a radioprotective chemical, injected i.p. into mice prior to either radiation quality alone or into combined injured mice increased 30-day survival and reduced susceptibility to challenge with Klebsiella pneumoniae. Protection against irradiation and resistance to bacterial challenge afforded by the WR compound was greater for γ-photon irradiation than for mixed-field irradiation. The TRIGA reactor can be used to simulate nuclear radiation-induced situations that include traumas or infections. Countermeasures for increasing survival after mixed-field irradiation may be more difficult than for γ-photon irradiated casualties. (authors)

  6. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  7. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  8. Reactor water supplementing facility

    International Nuclear Information System (INIS)

    Condensates stored in a main condenser are introduced to a turbine-driven reactor feed water pump by way of a low pressure condensate pump, a condensate cleanup facility, a high pressure condensate pump and a low pressure feed water heater by condensate pipelines. The turbine driven feed water pump introduces feed water by way of a high pressure feed water heater to a reactor pressure vessel (RPV). Further, an auxiliary condensate pipeline having a booster pump and connected at one end to the main condenser is connected to the upstream of a motor-driven reactor feed water pump. Downstream of the turbine-driven feed water pump is connected to the downstream of the electromotive feed water pump. Then, when the condensate pump or a turbine-driven feed water pump should stop and if start of a stand-by pump is failed due to some or other reason, the motor-driven feed water pump and the booster pump are started based on a pump stop signal. With such procedures, coolants are supplied to RPV thereby enabling to ensure coolant level in the RPV. (I.N.)

  9. Fast reactor fuel cycle facility

    International Nuclear Information System (INIS)

    An integrated fuel cycle facility named Fast Reactor Fuel Cycle Facility (FRFCF) is planned to be set up at Kalpakkam to close the fuel cycle of the Prototype Fast Breeder Reactor (PFBR) that is already under construction there. FRFCF is the first project of its kind in India. Closure of fuel cycle of PFBR will be a significant milestone of the second stage of nuclear power programme of the Department of Atomic Energy. The facility would be ready for operation in 2014. Design work and safety review of FRFCF are presently in progress. (author)

  10. Reactor water injection facility

    International Nuclear Information System (INIS)

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  11. Hot Fuel Examination Facility's neutron radiography reactor

    International Nuclear Information System (INIS)

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell

  12. Refurbishment status on reactor facilities of JMTR

    International Nuclear Information System (INIS)

    The JMTR (Japan Materials Testing Reactor), a light-water-cooling tank-type reactor with a 50 MW thermal power, was shutdown in August 2006. The reactor facilities are to be refurbished during four years from the beginning of FY 2007, and the renewed JMTR will restart from FY 2011. In advance of the reactor refurbishment, equipments on reactor facilities to be renewed and to be continuously used were selected from a viewpoint of ensuring safety, improvement of operating availability, etc. The selected equipments to be renewed were the reactor instrument and control system, cooling system, radioactive waste facility, power supply system, boiler, etc. This report describes the basic idea on selection of the renewal facilities and schedule of refurbishment work. (author)

  13. Test Facility for SMART Reactor Flow Distribution

    International Nuclear Information System (INIS)

    A Reactor Flow Distribution Test Facilities for SMART, named SCOP (SMART Core Flow and Pressure Test Facility), were designed in order to simulate the distributions of (1) core flow and (2) reactor sectional flow resistance and flow rates. SCOP facility was designed based on the linear scaling law in order to preserve the flow characteristics of the prototype system, which are distributions of flow rate and pressure drop. The reduced scale was selected as a 1/5 of prototype length scale. The nominal flow condition was designed to be similar based on the velocity as that of the SMART reactor, which can minimize the flow distortion in the reduced scale of test facility by maintaining high Re number flow. Test facility includes fluid system, control/instrumentation system, data acquisition system, power system, which were designed to meet the requirement for each system. This report describes the details of the scaling and design features for the test facility

  14. Facility modernization Annular Core Research Reactor

    International Nuclear Information System (INIS)

    The Annular Core Research Reactor (ACRR) has undergone numerous modifications since its conception in response to program needs. The original reactor fuel, which was special U-ZrH TRIGA fuel designed primarily for pulsing, has been replaced with a higher pulsing capacity BeO fuel. Other advanced operating modes which use this increased capability, in addition to the pulse and steady state, have been incorporated to tailor power histories and fluences to the experiments. Various experimental facilities have been developed that range from a radiography facility to a 50 cm diameter External Fuel Ring Cavity (FREC) using 180 of the original ZrH fuel elements. Currently a digital reactor console is being produced with GA, which will give enhanced monitoring capabilities of the reactor parameters while leaving the safety-related shutdown functions with analog technology. (author)

  15. Claim prevention at reactor facilities

    International Nuclear Information System (INIS)

    Why does a radiation worker bring a claim alleging bodily injury from radiation exposure? Natural cancer, fear of radiation induced cancer, financial gain, emotional distress and mental anguish are some reasons for workers' claims. In this paper the author describes what power reactor health physicists are doing to reduce the likelihood of claims by establishing programs which provide sound protection of workers, prevent radiological events, improve workers' knowledge of radiological conditions and provide guidance for radiological incident response

  16. ETRR-2 nuclear reactor: Facility specification

    International Nuclear Information System (INIS)

    The report provides technical details on the ETRR-2 nuclear reactor core and immediate structure for analysis purposes. The goal of the report is to provide sufficient geometric and material data to build a computational neutronic model of the facility. (author)

  17. Low-temperature irradiation facilities in reactors

    International Nuclear Information System (INIS)

    The general principle of low-temperature irradiation facilities in reactors is described and the main fields of research which are relevant to them are discussed. Existing facilities using He, H2, Ne or N2 as cooling liquids (or gases) are compared and some special problems as the production of nuclear heat, the choice of the cooling agent, safety considerations and continuous local flux monitors are spelled out. Finally, the liquid He facility in Garching is described along with some of the experimental techniques used and experiments performed. (K.B.)

  18. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  19. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  20. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    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)

  1. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  2. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  3. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  4. Risk management activities at the DOE Class A reactor facilities

    International Nuclear Information System (INIS)

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  5. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  6. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  7. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  8. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Science.gov (United States)

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. PMID:22154389

  9. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  10. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  11. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor's Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced

  12. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  13. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  14. Research Reactor Benchmarking Database: Facility Specification and Experimental Data

    International Nuclear Information System (INIS)

    This web publication contains the facility specifications, experiment descriptions, and corresponding experimental data for nine different research reactors covering a wide range of research reactor types, power levels and experimental configurations. Each data set was prepared in order to serve as a stand-alone resource of well documented experimental data, which can subsequently be used in benchmarking and validation of the neutronic and thermal-hydraulic computational methods and tools employed for improved utilization, operation and safety analysis of research reactors

  15. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  16. Experimental capabilities of the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service

  17. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  18. Design requirements for new nuclear reactor facilities in Canada

    International Nuclear Information System (INIS)

    The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

  19. Practical nuclear power training for overseas trainees using reactor facilities and radiation handling facilities

    International Nuclear Information System (INIS)

    The research reactor of Tokyo City University Atomic Energy Research Laboratory (Musashi Institute of Technology reactor) is zirconium-moderated water-cooled solid homogeneous type (TRIGA-II type), and its maximum heat output is 100 kW. It got into the first critical state in January 1963, and since then, it has achieved success in many researches. Although its decommissioning was decided in 2013, the existing facilities are used in education, and the research related to the decommissioning of research reactor facilities is carried out. Radiation handling facilities are in place, and they are widely used in education and research activities. Atomic Energy Research Laboratory, as a place for education, is conducting education and research activities such as the education using radiation handling facilities, development of an actual feeling type reactor operation simulator using the control panel of Musashi Institute of Technology reactor and operation performance data. This paper reports the practical nuclear power training for overseas trainees using the reactor facilities and radiation handling facilities. It also reports training implementation plan, acceptance preparation, contents of training, and the results of training. (A.O.)

  20. The TRIGA reactor Frankfurt construction and experimental facilities

    International Nuclear Information System (INIS)

    The new reactor FRF 2 was designed by Gutehoffnungshutte Sterkrade AG in cooperation with the reactor group of the Institut fur Kernphysik. The maximum power level is 1 MW; later installation of facilities for pulsed operation is possible. Performance and design data of the FRF 2 are given. The reactor is expected to start operation in 1973. Since the FRF 2 will be installed inside the biological shield and reflector of the FRF 1, the FRF 2 core has to correspond to the FRF 1 core structure

  1. SANS facility at the PITESTI 14MW TRIGA reactor

    International Nuclear Information System (INIS)

    At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest

  2. Present status of decommissioning in the Musashi Reactor Facility

    International Nuclear Information System (INIS)

    The Musashi reactor facility (TRIGA-II type) had been used as the place of studies such as the education and training of nuclear engineers, research on an activation analysis or reactor physics and a university joint research use mainly on medical and bio therapy study since the first criticality in 1963. Unfortunately a coolant leak was found from a core tank in December 1989 and stopped the operation of the reactor since. The Musashi reactor decommissioning was decided in May 2003 and took a step to shut down the reactor permanently. In 2006 all the spent fuels were returned to U.S.A. Outlines of the transport works were presented as present status of the decommissioning. (T. Tanaka)

  3. A neutronradiography facility based on an experimental reactor

    OpenAIRE

    THOMAS DIMITRIOS; J. G. Fantidis; NICOLAOU G.

    2014-01-01

    A thermal Neutron Radiography (NR) facility based on the use of thermal neutron flux, generated by the PULSTAR experimental reactor, has been designed and simulated using the MCNPX code. The key objective of the proposed facility is to deliver thermal neutron flux in this range for variable values of L/D ratio, instantaneously with acceptable values for all NR parameters. Thus, with suitable aperture and collimators designs, optimization for the parameters for thermal NR was achieved, for a w...

  4. Liquid Metal Fast Breeder Reactor Program (LMFBR): facility profiles

    International Nuclear Information System (INIS)

    A description is presented of the experimental test facilities involved in the conduct of the LMFBR research and development program. Existing facilities and those under construction or authorized as of October 1975 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. Introductory material for each section includes site and facility maps and an alphabetical list of the profiles contained in the section. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc. involved in the LMFBR program is included. Alphabetical, organizational, and programmatic indexes are provided as a convenient method of identifying the facilities with their locations and with their principal uses in the LMFBR program

  5. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U3Si2) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×1014 cm-2s-1, and is controlled by a Hafnium control absorber rod and B4C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  6. Cold neutron irradiation facility for the Brazilian research reactors

    International Nuclear Information System (INIS)

    Neutron irradiation in research reactors and accelerators can be realized at appropriated neutron guides or beam holes shared around a cold neutron source (CNS) with neutron of variable intensity and energy. An irradiation facility for multiple applications including an intense CNS was calculated for the three Brazilian research reactors and can be utilized as a first concept for the new research reactor to be designed, the Brazilian multiple purpose research reactor (RMB). A study about coolant and moderators properties, and simulations with neutron physics and thermal codes, may be important for the definition of the type of the CNS to be utilized. Some earlier results of MCNP simulations and a discussion about the different factors involved in the definition of its installation in the Brazilian research reactors are here presented. One suggests an international cooperation for the design development of this system and posterior construction of a prototype in the Argonauta reactor at the Instituto de Engenharia Nuclear (IEN-CNEN/RJ). It is also being considered the inclusion of other devices as a neutron fiber to guide the neutron beams away of the gamma radiation and fast neutron background. The cold neutron facility increases the intensity of cold neutrons, without the need of additional fuel burn up. (author)

  7. Guidance of clearance related standards in reactor facilities

    International Nuclear Information System (INIS)

    The reactor regulation law was amended in May 2005 to provide for the clearance system. The regulatory body confirmed radioactivity concentration of material and allowed the material be classified as 'material cleared from regulatory control'. The procedure of confirmation consisted of two steps; 1) methodology and 2) results, of measurement and assessment. Clearance related standards were issued as ordinance of Ministry and NISA information notice, which specified radioactive nuclides, radioactivity concentration, evaluation unit, how to decide radioactivity concentration, radiation measurement equipment and control of object material. This report was guidance of clearance related standards in reactor facilities with explanatory notes. By applying the clearance system, material of insignificant radiation level, which was part of concrete or metals generated from decommissioning of reactor facilities, could be released from regulatory control of radioactive waste as clearance material, and recycle of such material could result in reduction of radioactive waste amount. (T. Tanaka)

  8. The Reactor and Cold Neutron Research Facility at NIST

    International Nuclear Information System (INIS)

    The NIST Reactor (NBSR) and Cold Neutron Research Facility (CNRF) are located at the Gaithersburg, MD site, and have been in operation since 1969 and 1991, respectively. A total of 26 thermal neutron facilities and 11 cold neutron stations are operating for studies in condensed matter physics and chemistry, materials science, chemical analysis, nondestructive evaluation, neutron standards, fundamental neutron physics, and irradiations. Thermal and cold neutron instruments which have become operational since the 2d IGORR Conference will be described. Major facility upgrades to be implemented in early 1994 will be outlined. (author)

  9. Advanced Test Reactor National Scientific User Facility Partnerships

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of

  10. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  11. Environmental assessment for the deactivation of the N Reactor facilities

    International Nuclear Information System (INIS)

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

  12. Ventilation safety of facilities comprising nuclear reactors

    International Nuclear Information System (INIS)

    The reliability of the ventilation is one of the most important aspects in the prevention of the nuisances that a nuclear installation can provide, since the ventilation is located at the last barrier. A certain number of essential points have been recalled here. But it is necessary to bear in mind other requirements such as the limitation in the number of crossovers, the answers to be found should the system fail, the need to show that ventilation systems do not in themselves bring other nuisances such as noise, irradiation or contamination hazards, likelyhood of recycling the contamination, vibrations, fire. Finally, it is absolutely essential, right from the project stage, that the design ensures that very good accessibility, very easy dismantling and handling, as well as all the facilities needed to make sure of the initial and periodic tests, are guaranteed

  13. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  14. Fast reactor experiments with thorium at the PROTEUS facility

    International Nuclear Information System (INIS)

    The largescale utilization of thorium is usually linked to its introduction in fast breeder reactors and/or advanced converters. The present experiments were carried out in the zero-energy reactor facility, PROTEUS, at EIR. Six different configurations for the central fast test zone were considered in the current programme, the principal fuel/blanket materials used being in the form of rods of 15% PuO2/UO2, depleted UO2, ThO2 and Th-metal. For each configuration, measurements of the principal reaction rate ratios at the centre, as well as of reaction rate distributions across the test zone, were made. (Auth.)

  15. The Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

    2011-08-01

    In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those researchers with the best ideas access to the most advanced test capability, regardless of the proposer’s physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these university capabilities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user’s technical needs. The current NSUF partners are

  16. The PLUTO reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Technical information is given on the PLUTO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities and specialized irradiation devices (loops and capsules). The information is presented in the form of five information sheets under the headings; main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices

  17. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 1014 n/cm2/sec and a liquid D2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  18. The Advanced Test Reactor as a National Scientific User Facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR) has been in operation since 1967 and mainly used to support U.S. Department of Energy (US DOE) materials and fuels research programs. Irradiation capabilities of the ATR and post-irradiation examination capabilities of the Idaho National Laboratory (INL) were generally not being utilized by universities and other potential users due largely to a prohibitive pricing structure. While materials and fuels testing programs using the ATR continue to be needed for US DOE programs such as the Advanced Fuel Cycle Initiative and Next Generation Nuclear Plant, US DOE recognized there was a national need to make these capabilities available to a broader user base. In April 2007, the U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). As a NSUF, most of the services associated with university experiment irradiation and post-irradiation examinations are provided free-of-charge. The US DOE is providing these services to support U.S. leadership in nuclear science, technology, and education and to encourage active university/industry/laboratory collaboration. The first full year of implementing the user facility concept was 2008 and it was a very successful year. The first university experiment pilot project was developed in collaboration with the University of Wisconsin and began irradiation in the ATR in 2008. Lessons learned from this pilot program will be applied to future NSUF projects. Five other university experiments were also competitively selected in March 2008 from the initial solicitation for proposals. The NSUF now has a continually open process where universities can submit proposals as they are ready. Plans are to invest in new and upgraded capabilities at the ATR, post-irradiation examination capabilities at the INL, and in a new experiment assembly facility to further support the implementation of the user facility concept. Through a newly created Partnership Program

  19. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D2O/H2O ice cold source. (author)

  20. Nuclear technologies at WWR-K reactor facility

    International Nuclear Information System (INIS)

    The WWR-K nuclear research reactor was put into operation for the first time in November, 1967. It was put into operation again, after the 10-year outage, in March, 1998. Industrial application of the nuclear and radiation technologies in many economical fields of the country, particularly in establishment of reactor radioisotope production for the medicine, industry and scientific research, is of a great importance and promising for Kazakstan. The WWR-K reactor is practically the only suitable facility in the country, according to its technical, operating, neutron and physical performance, that may be used for establishment of radioisotopes production and development of such nuclear and radiation technologies as the neutron and neutron-capture treatment of the malignant tumours, sterilization of medical instruments and materials, nuclear doping of semi-conductors, neutron and activation analysis and others

  1. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  2. Reactor Simulator Development Facility for Operating Personnel Training

    International Nuclear Information System (INIS)

    A reactor simulator development facility (FARSim) for operator training is presented. This facility is a software product that can be divided into four main modules: the model manager (MM), the simulator human machine interface (SHMI), the instructor station (IS) and the simulation manager (SM). It is designed as a distributed system where each module takes charge of a specific simulator task that could run in the same computer or distributed in a computer network. The main module is the SM which is responsible for routing the messages between the other modules and managing the simulation. The MM interfaces to the plant mathematical model (PMM). The IS is the process where the instructor commands the simulation, performing tasks such as start, pause and stop the simulation. The SHMI is the interface with the simulator SCADA (which can be identical to the plant SCADA) and is used by the trainee to observe the simulated plant output and to act upon it. The PMM is built using Matlab-Simulink simulation engine and graphical design user interface, for which specific libraries have been developed with a comprehensive set of nuclear and thermohydraulic plant components. This simulator development facility is being used to develop the ANSTO replacement research reactor full-scope and partial replica reactor training simulator (RTS). (author)

  3. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  4. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and defuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  5. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  6. Overview of irradiation facilities and experiments currently in the Oak Ridge High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The Oak Ridge High Flux Isotope Reactor (HFIR) is an 85 MW research reactor with a variety of irradiation facilities. The target region has the highest continuous thermal neutron flux available in the western world and facilities in the beryllium reflector provide opportunities to irradiate experiments of various sizes in a variety of neutron spectrums. Major programs utilizing these facilities include Fusion Materials, Advanced Neutron Source (ANS), New Production Reactor, and Modular High Temperature Gas-Cooled Reactor

  7. 1996 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  8. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    International Nuclear Information System (INIS)

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830

  9. The new neutron imaging facility at TRIGA reactor in Morocco

    International Nuclear Information System (INIS)

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.1013ncm2/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  10. Research reactor and fuel development facility decommissioning experience and technology

    International Nuclear Information System (INIS)

    This paper discusses the technology and experience gained in research reactor and fuels development facility decommissioning programs carried out by Babcock and Wilcox (B and W) at one of its NRC-licensed sites in Lynchburg, VA. The projects included two buildings that housed plutonium/uranium fuels development laboratories, four low-power critical experiment facilities, and two (megawatt-level) research reactors. This paper concentrates on the experiences with the plutonium/uranium fuels development laboratories and critical experiment facilities. These were comprehensive projects that included: developing the decommissioning and quality assurance plans; interfacing with the U.S. Nuclear Regulatory Commission, performing the actual decontamination/dismantling work; performing decontamination and final radiological surveys; and volume reducing, packaging, certifying, classifying, and shipping the radioactive waste for disposal. This broad experience has involved handling radioactive contamination from the following sources: low- and high-enriched U-235 fuel; depleted uranium; mixed oxide fuel (Pu/UO); thorium fuel; U Al alloy fuel; and fission activation products (beta-gamma). Areas of application to future projects are highlighted in this paper

  11. The new neutron imaging facility at TRIGA reactor in Morocco

    Energy Technology Data Exchange (ETDEWEB)

    Ouardi, A.; Alami, R.; Bensitel, A. [Centre National de l' Energie des Science et des Techniques Nucleaires, PB.1382 R.P 10001 Rabat (Morocco)

    2011-07-01

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.10{sup 13}ncm{sup 2}/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  12. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.)

  13. Personal neutron dosimetry at a research reactor facility

    International Nuclear Information System (INIS)

    Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

  14. Reactor safety research program at Thai test facility

    International Nuclear Information System (INIS)

    Thermal-hydraulics, Hydrogen, Aerosol and Iodine (Thai) aims at providing experimental database for the verification and validation of Lumped Parameter (Lp) and Computational Fluid Dynamics (CFD) codes with 3-dimensional capabilities. Since its construction in 2000, Thai facility has been engaged in the field of reactor safety in the frame of various national (Thai I: 2000-2003, Thai II: 2003-2006, Thai III: 2006-2009, Thai IV: 2009-2012) and international programs (OECD-Thai: 2007-2009). Additionally, experimental data has been provided for several international standard problems (ISP 41, 46, 47 and 49) code validation exercises. Experiments performed in Thai facility cover a wide spectrum or reactor safety relevant issues by investigating separate and coupled-phenomenon experiments under design basis accident and severe-accident-typical scenarios. Experiments are performed in close co-operation with AREVA Erlangen and Grs Koln. Experimental configuration and the operating conditions in Thai vessel typical of those for PWR, BWR and High Temperature Gas Cooled Reactor can be produced thanks to its modular structure, appropriate feeding/generation devices for gases (H2, He, Steam, N2, etc.), Aerosol (inert and hygroscopic), Iodine Radiotracer, and advanced instrumentation. Experiments also cover investigation of passive safety systems, e.g. commercial Par for H2 mitigation in phenomenon orientated experiments to enhance the confidence in the performance of passive mitigation systems during severe accident scenarios and also to establish a common database accessible by a large research community to support further development and validation of the Lp and CFD codes with 3-dimensional capabilities. This paper summarizes experimental investigations made in Thai test facility to investigate issues related to the thermal-hydraulics, fission product (aerosol, iodine) transport and their interaction with containment walls (deposition, resuspension) and passive safety

  15. Neutron radiography and tomography facility at IBR-2 reactor

    Science.gov (United States)

    Kozlenko, D. P.; Kichanov, S. E.; Lukin, E. V.; Rutkauskas, A. V.; Belushkin, A. V.; Bokuchava, G. D.; Savenko, B. N.

    2016-05-01

    An experimental station for investigations using neutron radiography and tomography was developed at the upgraded high-flux pulsed IBR-2 reactor. The 20 × 20 cm neutron beam is formed by the system of collimators with the characteristic parameter L/D varying from 200 to 2000. The detector system is based on a 6LiF/ZnS scintillation screen; images are recorded using a high-sensitivity video camera based on the high-resolution CCD matrix. The results of the first neutron radiography and tomography experiments at the developed facility are presented.

  16. Dhruva reactor -- a high flux facility for neutron beam research

    International Nuclear Information System (INIS)

    Dhruva reactor, the highest flux thermal neutron source in India has been operating at full power of 100 MW over the past two years. Several advanced facilities like the cold source, guides, etc. are being installed for neutron beam research in condensed matter. A large number and variety of neutron spectrometers are operational. This paper deals with the basic advantages that one can derive from neutron scattering investigations and gives a brief description of the instruments that are developed and commissioned at Dhruva for neutron beam research. (author). 3 figs

  17. The SANS facility at the Pitesti 14MW TRIGA reactor

    International Nuclear Information System (INIS)

    The SANS facility existing at the Pitesti 14MW TRIGA reactor is presented. The main characteristics and the preliminary evaluation of the installation performances are given. A monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A is produced by a mechanical velocity selector with helical slots. A fruitful partnership was established between INR Pitesti (Romania) and JINR Dubna (Russia). The first step in this cooperation consists in the manufacturing in Dubna of a battery of gas-filled positional detectors devoted to the SANS instrument

  18. The neutron radiography facility at Tehran Research Reactor (TRR)

    International Nuclear Information System (INIS)

    Full text: Non-destructive testing in many fields of industry including detection of explosives, at the airports, testing for micro-cracks on airplane wings and turbine blades cracks is badly needed. Thermal neutron beam is one of preferable method to detect the micro-cracks, reveals the internal structure of components and explosives. The purpose of this paper is to present the neutron radiography facility at Tehran Research Reactor (TRR), Science and Technology Research Institute, and in particular to emphasize the industrial applications in wood industry, automobile engine inspection, minerals composition identification, turbine blade cracks detection. (author)

  19. A neutronradiography facility based on an experimental reactor

    Directory of Open Access Journals (Sweden)

    D. T. Thomas

    2015-06-01

    Full Text Available A thermal Neutron Radiography (NR facility based on the use of thermal neutron flux, generated by the PULSTAR experimental reactor, has been designed and simulated using the MCNPX code. The key objective of the proposed facility is to deliver thermal neutron flux in this range for variable values of L/D ratio, instantaneously with acceptable values for all NR parameters. Thus, with suitable aperture and collimators designs, optimization for the parameters for thermal NR was achieved, for a wide range of the collimator ratio. The short time requirements for obtaining the radiography images justify the use of the proposed system for ‘real time radiography’. The system was designed under the limitation that the total Dose Equivalent Rate does not exceed at the external shield surface the limit recommended by ICRP-26.

  20. Neutron diffraction facilities at the high flux reactor, Petten

    Science.gov (United States)

    Ohms, C.; Youtsos, A. G.; Bontenbal, A.; Mulder, F. M.

    2000-03-01

    The High Flux Reactor in Petten is equipped with twelve beam tubes for the extraction of thermal neutrons for applications in materials and medical science. Beam tubes HB4 and HB5 are equipped with diffractometers for residual stress and powder investigations. Recently at HB4 the Large Component Neutron Diffraction Facility has been installed. It is a unique facility with respect to its capability of handling heavy components up to 1000 kg in residual stress testing. Its basic features are described and the first applications on thick piping welds are shown. The diffractometer at HB5 can be set up for powder and stress measurements. Recent applications include temperature dependent measurements on phase transitions in intermetallic compounds and on Li ion energy storage materials.

  1. Advanced Test Reactor National Scientific User Facility Progress

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  2. SANS facility at the Pitesti 14 MW Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A. [Institute for Nuclear Research - Pitesti (Romania); Grabcev, B.; Todireanu, S. [National Institute of Materials Physics (NIMP) Bucharest (Romania); Constantin, F. [National Institute of Physics and Nuclear Engineering Bucharest (Romania); Shvetsov, V. [Joint Institute for Nuclear Research Dubna (Russian Federation); Popescu, G. [National College Al. Odobescu Pitesti (Romania)

    2006-07-01

    Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom {<=} {lambda} {<=} 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Q{sub min} = 0.005 Angstrom{sup -1} while the maximal value is Q{sub max} = 0.5 Angstrom{sup -1}. The relative error is {delta}Q/Q{sub min} = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia

  3. SANS facility at the Pitesti 14 MW Triga reactor

    International Nuclear Information System (INIS)

    Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom ≤ λ ≤ 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Qmin = 0.005 Angstrom-1 while the maximal value is Qmax = 0.5 Angstrom-1. The relative error is ΔQ/Qmin = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation consists

  4. SANS Facility at the Pitesti 14 MW TRIGA Reactor

    International Nuclear Information System (INIS)

    At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nanocrystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the TRIGA Reactor of the Institute of Nuclear Research in Pitesti, Romania becomes actual especially after the shutting down of the WWR-S Reactor from Bucharest. A monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A is produced by a mechanical velocity selector with helical slots. The distance between sample and detectors plane is 5.2 m. The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Qmin = 0.005 A-1 while the maximal value is Qmax = 0.5 A-1. The relative error is ΔQ/Qmin = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation consists in the manufacturing at Dubna of a battery of gas filled

  5. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    International Nuclear Information System (INIS)

    The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy

  6. 1997 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  7. Naval Reactors Facility environmental monitoring report, calendar year 2001

    International Nuclear Information System (INIS)

    The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy

  8. 1993 environmental monitoring report for the naval reactors facility

    International Nuclear Information System (INIS)

    The results of the radiological and nonradiological environmental monitoring programs for 1993 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  9. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    International Nuclear Information System (INIS)

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  10. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    International Nuclear Information System (INIS)

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  11. 1991 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    The results of the radiological and non-radiological environmental monitoring programs for 1991 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were within the guidelines established by state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or heath and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the EnVironmental Protection Agency (EPA) and the Department of Energy (DOE)

  12. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  13. Thermal facility for BNCT in RA-1 Argentine research reactor

    International Nuclear Information System (INIS)

    Full text: A thermal facility for BNCT experiments is being developed in an Argentine Research Reactor: RA-1 'Enrico Fermi'. RA-1 research nuclear reactor is working at Constituyentes Atomic Center, near Buenos Aires, and started operations in 1958. It worked at several power levels, up to 120 k W. Today, RA-1 is licensed to work at 40 k W. RA-1 was used to produce radioisotopes in the early 60's, and today gives irradiation services to test materials, to calibrate detectors and activation analysis. RA-1 users are CNEA researchers, Nuclear Regulatory Authority staff and private laboratories. Boron Neutron Capture Therapy (BNCT) is a method to fight against cancer. It consists to irradiate cancer tumors using thermal neutrons. The tumor tissue should include a dose of a boron solution. The Boron irradiation produces the following nuclear reactions: n + B10→ α + Li7 + γ. Being the α particle a radiation with short range, but high destructive energy, the tumor cells are destroyed. The neutron flux should be of 109 n/cm2seg, and the gamma dose lower than 0.48 s V/h. This method is oriented to treat brain tumors. Taking in account that the brain tumors usually are several centimeters deep in the head, to get thermal neutrons in the tumor is convenient to irradiate the patient using epithermal neutrons. moderation in the cells of the brain will permit to get more thermal neutrons in the tumor. In CNEA BNCT program there is in construction an epithermal clinical facility in the RA-6, a 500 k W research reactor that is at Bariloche Atomic Center. To perform some experiments for instance to test the boron compounds, RA-1 is used. In this experiments little animals like hamsters or bottles with cultivated cells are used, for that reasons thermal neutrons are used. The project in RA-1 consists in several stages. As the first stage a preliminary thermal facility was built. Irradiation times of 45-60 minutes were estimated, at power operation levels of 40 k W. Several

  14. The Cairo Fourier Diffractometer Facility At The ETRR-1 Reactor

    International Nuclear Information System (INIS)

    The work presents the Cairo Fourier diffractometer facility (CFDF). The CFDF is based on the reverse time-of-flight (RTOF) concept and was recently installed, as IAEA TC Project, at one of the ET-RR-1 reactor horizontal channels. The CFDF performance is assessed and its main parameters are given. The facility applies a Fourier chopper system and 6Li-glass scintillators (NE-912) arranged according to the time focusing geometry in order to detect neutrons scattered from the sample at an angle 29= 90 diameter. The detector system has been optimized for studying the internal stresses in materials along with neutron diffraction measurements. Its angular aperture was found, from precise calculations, by a special program, to be equal to S.1 x 102 steradians. The neutron guide system attached to the CFDF provides a thermal neutron flux-l.lxl06n/cm2/sec at the sample position. It has been found, from measurements with different powder samples, that such value of the thermal neutron flux is adequate for neutron diffraction measurements, at scattering angle 2θ= 90ο and d-spacing values between 0.7 A and 2.5 A, within 0.45% resolution

  15. Post-irradiation examination of Fugen reactor fuel assembly at reactor fuel examination facility

    International Nuclear Information System (INIS)

    Post-irradiation examination of the first assembly of a monitoring program for Heavy Water Reactor ''Fugen'' of PNC (Power Reactor and Nuclear Fuel Development Corporation) has been executed since Oct. 1983 at the Reactor Fuel Examination Facility, JAERI Tokai (Japan Atomic Energy Research Institute, Tokai Research Establishment). The fuel assembly is a cylindrical cluster, with 4,400mm length, composed of 28 rods in 3 concentric circles, 12 spring-grid spacers and the upper and lower tie plates. The fuel is plutonium-uranium mixed oxide (0.8 w/o), and the material of cladding tube is Zry-2. The average burnup of the fuel assembly is about 13,600 MWd/t. This paper describes the methods and some results on the post irradiation examination items as follows: 1. Radioactive measurement of water in transportation cask; 2. Visual inspection of the fuel assembly in dry cell, before and after removing the crud, by ultrasonic vibration method; 3. Chemical analyses and radioactive measurement of the crud materials; 4. Dimensional measurement of assembly length and rod-rod gaps, before and after removing the crud; 5. Disassembly and dimensional measurement of rod-rod gaps in the inner circles; 6. Several nondestructive testing techniques of fuel rods. (author)

  16. The BNCT facility at the HFR Petten: Quality assurance for reactor facilities in clinical trials

    International Nuclear Information System (INIS)

    The first clinical trial in Europe of Boron Neutron Capture Therapy (BNCT) for the treatment of glioblastoma was opened in July 1997. The trial is a Phase I study with the principal aim to establish the maximum tolerated radiation dose and the dose limiting toxicity under defined conditions. It is the first time that a clinical application could be realised on a completely multi-national scale. The treatment takes place at the High Flux Reactor (HFR) in Petten, the Netherlands, is operated by an international team of experts under the leadership of a German radiotherapist, and treats patients coming from different European countries. It has therefore been necessary to create a very specialised organisation and contractual structure with the support of administrations from different countries, who had to find and adapt solutions within existing laws that had never foreseen such a situation. Furthermore, the treatment does not take place in an hospital environment and even more so, the facility is at a nuclear research reactor. Hence, special efforts were made on quality assurance, in order that the set-up at the facility and the personnel involved complied, as closely as possible, with similar practices in conventional radiotherapy departments. (author)

  17. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  18. Reactor physics experiments with thorium based clusters in AHWR - critical facility

    International Nuclear Information System (INIS)

    AHWR - Critical Facility (AHWR - CF) is a 'zero power' reactor designed to carry out various reactor physics experiments for validation of AHWR design. A number of experiments have been carried out in standard and extended reference core of the reactor. In this paper, results of experiments with different Thorium based experimental clusters are presented. These experiments provided valuable data for validation of reactor physics design methodologies. (author)

  19. Reactor simulator development facility for operating personnel training

    International Nuclear Information System (INIS)

    Full text: A development facility of reactor simulators (FARSim) for operator training is presented. The simulator development facility can be divided into four main modules: the model manager (MM), the simulator human machine interface (SHMI), the instructor station (IS) and the simulation manager (SM). It is designed as a distributed system where each module takes charge of an specific simulator task that could run in the same computer or distributed in a computer network. Only the SHMI runs in a remote computer (specific hardware). The process distribution is configurable at the start of the simulation session. This type of process distribution makes it scalable. The main module is the Simulation Manager (SM) which is responsible of routing the messages between the other modules and managing the simulation. The Model Manager (MM) interfaces to the plant mathematical model (PMM). The Data Base Manager (DBM) handles the data base, in order to access and save necessary information during the simulation. The timing of the simulation is accomplished by the Clock Manager (CM) and the error and system messages are handled by the Logger. The Instructor Console (IC) is the process where the instructor commands the simulation. The SHMI is a process interface with the simulator SCADA (which can be identical to the plant SCADA) and is used by the trainee to observe the simulated plant output and to act upon it. The PMM source code, together with the necessary libraries is encapsulated into the MM to implement the model initialisation and the one-step simulation. The MM provides the initial conditions (ICs), and the model input, the PMM provides the plant output. Depending on the complexity of the model, it can be divided and the calculations can be distributed among different CPUs. The PMM source code generation is based on the Matlab-Simulink-Real Time Workshop simulation development environment. The models are developed on graphical windows interfaces based on component

  20. Effect of potential energy stored in reactor facility coolant on NPP safety and economic parameters

    International Nuclear Information System (INIS)

    Potential (non-nuclear) energy stored in reactor facility coolant is a crucial factor determining the NPP safety/hazard characteristics as it is inherent property of the material and cannot be changed. Enhancing safety of the NPP with traditional type reactor facilities, in which potential energy is stored in large quantities, requires buildup of the number of safety systems and in-depth defense barriers, which reduce the probability of severe accidents (but do not exclude the opportunity of their realization) and seriousness of their consequences. Keeping the risk of radioactivity release for different type reactor facilities at a same level of social acceptability, the number of safety systems and in-depth defense barriers, which determine essentially the NPP economical parameters, can be reduced with diminishing the potential energy stored in the reactor facility. To analyze the effect of potential energy on reactor facility safety/hazard, a diagram of reactor facility hazard has been proposed. It presents a probability of radioactivity release as a function of radioactivity release values for reactor facilities with identical radiation potential, which differ by values of potential energy stored in coolant. It is proposed to account NPP safety/hazard effect on economics by adding a certain interest on the electricity cost for making payments in a special insurance fund assigned to compensate the expenses for elimination of consequences of a possible accident. (authors)

  1. Seismic response analyses for reactor facilities at Savannah River

    International Nuclear Information System (INIS)

    The reactor facilities at the Savannah River Plant (SRP) were designed during the 1950s. The original seismic criteria defining the input ground motion was 0.1 G with UBC provisions used to evaluate structural seismic loads. Later ground motion criteria have defined the free field seismic motion with a 0.2 G ZPA and various spectral shapes. The spectral shapes have included the Housner spectra (TID-7094), a site specific spectra, and the US NRC Reg. Guide 1.60 shape. The development of these free field seismic criteria are discussed in the paper. The more recent seismic analyses have been of the following type: fixed base response spectra, frequency independent lumped parameter soil/structure interaction (SSI), frequency dependent lumped parameter SSI, and current state of the art analyses using computer codes such as SASSI. The results from these computations consist of structural loads and floor response spectra (used for piping and equipment qualification). These results are compared in the paper and the methods used to validate the results are discussed

  2. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  3. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  4. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  5. A neutron tomography facility at a low power research reactor

    Science.gov (United States)

    Koerner, S.; Schillinger, B.; Vontobel, P.; Rauch, H.

    2001-09-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3×10 5 neutrons/cm 2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm.

  6. A neutron tomography facility at a low power research reactor

    International Nuclear Information System (INIS)

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3x105 neutrons/cm2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm

  7. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  8. A proposed irradiation-research facility to replace the NRU reactor

    International Nuclear Information System (INIS)

    This report describes the replacement of the National Research Universal (NRU) reactor with a dual purpose irradiation research facility to test CANDU fuels and materials, and to perform materials research using neutrons

  9. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  10. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  11. Verification of computer codes for dynamic processes in nuclear reactors against experiments at loop facility of IGR-1 pulse reactor

    International Nuclear Information System (INIS)

    Basic principles of PRISDG and PRISET computer codes structure to analyze dynamic processes in nuclear reactors are presented. The codes were verified against experimental studies of dynamic processes related with flow-stop and power surge. The experimental data were obtained at loop facility of IGR-1 pulse reactor using fuel assemblies of IVV-2M research reactor. Accuracy of the codes is the same as the accuracy achieved in the experiments. Analysis could be performed at PS-2-type personal computers. Running time is not longer than several tens of minutes. (author)

  12. Method of surface treatment for structure and facility in reactor

    International Nuclear Information System (INIS)

    Surfaces of weld zones, in contact with liquid, of structures and equipments in a reactor made of austenite stainless steels disposed in the reactor water of a reactor pressure vessel are melted by laser. Then, heat affected zones and grain boundary segregation portions, etc. with low corrosion resistance formed under irradiation are melted by laser beams and the molten surfaces are quenched by the surrounding reactor water. In this case, ferrites are formed to provide a two-phase structure. This can improve the corrosion resistance. Further, plasma technology can be used instead of the laser method. (I.S.)

  13. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  14. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  15. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    OpenAIRE

    Alessandro Alemberti; Fabio Berra; Davor Grgic; Graydon Yoder; Stefano Monti; Paride Meloni; Davide Papini; Fosco Bianchi; Marco Ricotti; Roberta Ferri; Cinzia Congiu; Gustavo Cattadori; Andrea Achilli; Bojan Petrovic; Andrea Maioli

    2009-01-01

    IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be bu...

  16. Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

    International Nuclear Information System (INIS)

    The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

  17. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits

  18. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  19. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  20. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  1. Over view of post-irradiation examination facilities for fuels and materials development of fast reactor

    International Nuclear Information System (INIS)

    The hot cell complex for post-irradiation examination of the fast reactor fuels and materials was constructed and has been operated at the Oarai engineering center of Japan Nuclear Cycle Development Institute. The complex consists of three hot cell facilities. They are the Fuel Monitoring Facility (FMF), the Alpha-Gamma Facility (AGF) and the Materials Monitoring Facility (MMF). The FMF is located adjacent to the experimental fast reactor 'JOYO' and started operation in November 1978. In this facility, nondestructive examination of fuel subassemblies and other core components, in addition to some destructive examination of fuel and absorber pins, are carried out. The selected pins and materials, sectioned to the appropriate size at the FMF, are sent to the AGF and the MMF for further detailed examinations. The AGF has been operated successfully since October 1971. The functions of this facility are the physical, metallurgical and chemical examinations of irradiated plutonium-bearing fuels. The MMF was constructed at 1972 and has been operated since 1973 for the reactor materials. In this facility, various tests are conducted on core materials, structural material and control rod materials irradiated in fast reactor. Structural materials irradiated in JMTR and pressure tubes irradiated in prototype advanced thermal reactor 'Fugen' are also examined. (author)

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

    International Nuclear Information System (INIS)

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

  3. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    OpenAIRE

    Guerra, Bruno T; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal, Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

  4. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... Opportunity for Hearing published in the Federal Register on August 19, 2011 (76 FR 52018- 52022). The NRC... Register on March 30, 2012 (77 FR 19362-19366), and concluded that renewal of the facility operating... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112...

  5. The ''CAMERA'' test facility in the OSIRIS reactor

    International Nuclear Information System (INIS)

    CAMERA is an irradiation installation conceived to measure under neutronic flux and continuously the dimension variations of a fuel pencil of PWR reactors. The device, set in the periphery of the OSIRIS reactor, can receive new, preirradiated or reconstituted pencils. The principles of measurements is explained. Then, a brief description of the installation is given: in-pile part; out-of-pile part; connections. The technical characteristics of the installation are presented. A first qualification test of the installation under flux has been carried out at the end of the first semester 1984 in the OSIRIS reactor

  6. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    International Nuclear Information System (INIS)

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MWt, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 1018 n·m-2·s-1 in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs

  7. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    International Nuclear Information System (INIS)

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MWt, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 1018 N·m-2·s-1 in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs

  8. The neutron radiography facility designed for TRIGA reactors and its results

    International Nuclear Information System (INIS)

    The two TRIGA reactors of INR, the Steady State Reactor (SSR) having a power of 14 MW and Annular Core Pulsing Reactor (ACPR) having in steady state a power of 500 kW and being capable of a pulse to the peak power of 20000 MW, are placed in the same pool. The neutron flux ranging at the edges of those reactors cores is suitable for neutron radiography. The neutron radiography facility is placed in the pool of the TRIGA reactors. Till now as neutron source only the ACPR, in steady state or pulsing mode has been used. For the future one intends to use also the neutron flux of SSR. The aim of this facility is to achieve neutron radiographs of the nuclear fuel elements. (authors)

  9. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.)

  10. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  11. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    International Nuclear Information System (INIS)

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam

  12. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Lee, C. S.; Seo, C. G

    2003-07-01

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam.

  13. Current development at the Finnish TRIGA reactor towards the operation of the new BNCT irradiation facility

    International Nuclear Information System (INIS)

    The FiR 1-reactor, a 250 kW Triga reactor, with its subsystems has experienced a large renovation work. The main purpose of the upgrading has been to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. During the renovation the ventilation, electricity lines, water and waste water pipe lines and the reactor cooling system were renewed. The epithermal beam facility of the BNCT-irradiation station got its final form. In order to increase the availability of the reactor and the necessary systems in the reactor building all the pumps of the reactor cooling system and the main fans of the ventilation have been doubled. The reactor instrumentation is fed by an Uninterruptible Power System and a diesel aggregate feeds power to the reactor cooling and the ventilation systems, if the normal power supply fails. The epithermal neutrons are produced from the fast fission neutrons by a moderator block consisting of Al+AIF3 (FLUENTALTM), Which showed to be the optimum material for this purpose. Independently of the large renovation work the application for a new operating license for the reactor had to be submitted at the end of the year 1998 after nine years' operating time With the old license. (author)

  14. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  15. The status of facilities at China Advanced Research Reactor

    International Nuclear Information System (INIS)

    A 60 MW research reactor, so called China Advanced Research Reactor (CARR,) was built in China Institute of Atomic Energy (CIAE), located in the southwest of Beijing and about 37 kilometers away from the central city. CARR is a tank-in-pool inverse neutron trap type reactor using D2O reflector, the designed optimal undisturbed thermal neutron flux is 8×1014 n⋅cm-2⋅s-1. A liquid D2 cold source will be equipped and the installation will be finished at the end of 2015. As a multipurpose research reactor, its main applications include neutron scattering, neutron activation analysis, isotope production, silicon doping, fuel element test, fundamental nuclear physics and so on. On March 13rd, 2012 CARR realized the 72 h stable operation with the full power. And the official operation license is expected to be issued at the beginning of next year. Cooperating with the internal and international users in the first phase ten instruments complete construction and are under commissioning, which are High Resolution Powder Diffractometer, High Intensity Powder Diffractometer, Residual Stress Diffractometer, Texture Diffractometer, Four Circle Diffractometer, Reflectometer, Small Angle Neutron Scattering, two Thermal Triple Axis Spectrometers and Isotope Separator On-Line instrument . In the second phase 7 instruments were approved and are under construction now. Although the operation license was not issued, the reactor was permitted to do the testing run several times and some results were obtained during the instrument commissioning.

  16. Development of facilities to irradiate materials in the RA1 and RA3 experimental reactors

    International Nuclear Information System (INIS)

    To study the properties of the materials under irradiation, devices and facilities were designed to work at experimental reactors of National Atomic Energy Commission. The radiological protection of the operators and the influence of the irradiated materials on the radiological inventory of the reactors were the most important aspects considered during the design stage. In the present work devices to operate in the argentine reactor 'Reactor Argentino (RA)', RA1 and RA3 experimental reactors are shown. These devices are dedicated to the study of the radiation damage by measuring property changes related to dimensional integrity and embrittlement of materials in zirconium alloys, steels and other materials used in nuclear reactors. The emphasis is on the previsions adopted to minimize the activation of their components and the criteria applied to guarantee the safety of the operators during their performance and after their subsequent dismantling. (author)

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

    International Nuclear Information System (INIS)

    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

  18. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  19. Conversion of the redundant research nuclear reactor into low power facility

    International Nuclear Information System (INIS)

    Most of reactors of the first generation have exhausted their operating life and should be dismantled. But nuclear reactor dismantling is a very complex process, full-scale realization of which generates a lot of other problems, solution of which is apparently beyond many countries strength. The goal of this report is to attract attention to the problem of decommissioning of the research nuclear reactor of the Institute of Physics of the Georgian Academy of Sciences, which has outlined its usefulness, and to search the acceptable ways of the nuclear reactor conversion into a state the maintenance of full radiation safety of which would not demand constant supervision and large expenditures. It is stated that the option of decommissioning of nuclear reactor IRT-M suggested by the Institute of Physics is: Radiation safe, ecologically clean and seismic stable; Comparably less labor-consuming and feasible; Not connected with the problem of formation of additional solid or liquid radioactive waste, and hence, with the problem of their accumulation and transportation; Not connected with large financial and material expenditures; Provides the possibility to give to the reactor a new function by installing a low power nuclear facility in the radiation-free part of the reactor tank for the purpose of Neutron Activation Analysis; Gives the possibility to use a new low power nuclear facility for attracting and training the young generation of specialists in the field of reactor physics and nuclear technology

  20. A proposal of reactor physics research of accelerator drive system using transmutation physics experimental Facility

    International Nuclear Information System (INIS)

    Reactor physics section of the Atomic Energy Society of Japan (AESJ) recognizes an accelerator driven system (ADS) as the next generation reactor and to promote researches using it. History of this section activity on ADS, outline of Transmutation Physics Experimental Facility in the 'High-Intensity Proton Accelerator Project', a proposal of reactor physics section to the project and future actions of this section are explained. The Transmutation Physics Experimental Facility consists of a fast neutron subcritical system and a nuclear spallation neutron source. The contents of experiments are evaluation of nuclear properties of fast neutron subcritical system driven by nuclear spallation source, verification of operation and control of accelerator driven hybrid system and evaluation of nuclear transmutation characteristics of MA (Minor Actinides) and LLFP (Long-Lived Fission Product). Themes of R and D of ADS contain operation control of ADS, critical control of subcritical system, properties of reactor with nuclear spallation neutron source and nuclear transmutation characteristics. The experimental items are measurement of dynamic characteristics of reactor at beam change, R and D of method of output control and stop, R and D of contentious monitoring method of subcritical multiplication, measurement of dynamic characteristics of behaviors of reactivity, effects on reactor characteristics of high energy neutron, effects on reactor physics of beam duct and large target, nuclear transmutation efficiency and simulation of nuclear transmutation reactor core. (S.Y.)

  1. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    Science.gov (United States)

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

  2. In-core dosimetry in CAGR - measurements on power reactors and laboratory facilities

    International Nuclear Information System (INIS)

    The problem of radiolytic corrosion of the graphite moderator in CAGR has led to a need for more accurate information on the radiation dose to the coolant gas in the pores of the graphite. An experimental in-core dosimetry programme is in progress to acquire this data. The problems of in-core dosimetry, particularly that of measuring gamma dose in the presence of high thermal neutron fluences, are described with reference to calorimetry, ionisation chambers and thermoluminescence dosimeters. Progress made in the refinement of these techniques for reactor dosimetry is described. An experiment is described in which dosimetry measurements in components of a Heysham Power Station reactor were made during its commissioning. The major facility of this dosimetry programme is a zero-energy research reactor constructed from CAGR components; this reactor and its experimental facilities are described, together with the results of some of the first experiments. (author)

  3. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  4. Extracted neutron beams experimental facilities and program of the first experiments at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Structural specific features of the IBR-2 pulse research biological hidered reactor. The characteristics of spectrometer for investigating the small angle neutron scattering and the CORA facility intended for investigating the structure and dynamics of condensed media by means of the therrol neutron scattering as well as the DN-2 diffractometer for investigating the atomic structure and crystallographic characteristics of monocrystals, having large (>10 A) elementary cell size and the equipment of the ultracold neutron channel are given. Biological shields of the reactor and experimental facilities are assembled of concrete blocks and standard building constructions and attains in the most dangerous regions 1 m. The shield ensures for the personnel a safety level of ionizing radiations and effectively shields the facilities from mutual effects caused by scattered radiation. The program of physical investigations planned at the IBR-2 continues the investigations started at the IBR-30 reactor

  5. Self-sustainability of a research reactor facility with neutron activation analysis

    International Nuclear Information System (INIS)

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  6. Simulation of the neutron flux in the irradiation facility at RA-3 reactor

    International Nuclear Information System (INIS)

    A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comisión Nacional de Energía Atómica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutron flux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility.

  7. The Irradiation Facility at the IBR-2M Reactor in Dubna

    Science.gov (United States)

    Bulavin, M.; Cheplakov, A.; Kulikov, S.; Kukhtin, V.; Shabalin, E.; Verhoglyadov, A.

    2014-06-01

    After renovation neutron pulsed reactor was successfully put into operation. In view of HL-LHC development, the irradiation facility at the IBR-2m, previously used for the ATLAS Liquid Argon calorimeters electronics and materials certification, has been upgraded to permit accumulation in the warm temperature conditions of the fast neutrons fluences up to k*1017 n/cm2 per one standard reactor operation cycle - 11 days. The first test of materials for ATLAS was conducted.

  8. The regulation and licensing of research reactors and associated facilities in the United Kingdom

    International Nuclear Information System (INIS)

    In the United Kingdom, the Nuclear Installations Inspectorate (NII) licenses nuclear facilities, including research reactors, on behalf of the Health and Safety Executive (HSE). The legislation, the regulatory organizations and the methods of operation that have been developed over the last 30 years result in a largely non-prescriptive form of control that is well suited to research reactors. The most important part of the regulatory system is the license and the attachment of conditions which it permits. These conditions require the licensee to prepare arrangements to control the safety of the facility. In doing so the licensee is encouraged to develop a 'safety culture' within its organization. This is particularly important for research reactors which may have limited staff resources and where the ability, and at times the need, to have access to the core is much greater than for nuclear power plants. Present day issues such as the ageing of nuclear facilities, public access to the rationale behind regulatory decisions, and the emergence of more stringent safety requirements, which include a need for quantified safety criteria, have been addressed by the NII. This paper explores the relevance of such issues to the regulation of research reactors. In particular, it discusses some of the factors associated with research reactors that should be considered in developing criteria for the tolerability of risk from these nuclear facilities. From a consideration of these factors, it is the authors' view that the range of tolerable risk to the public from the operation of new research reactors may be expected to be more stringent than similar criteria for new nuclear power plants, whereas the criteria for tolerable risk for research reactor workers are expected to be about the same as those for power reactor workers

  9. The BR1 research facilities to calibrate fuzzy logic technology for nuclear reactor control

    International Nuclear Information System (INIS)

    During the last three decades SCK-CEN has participated in various international programmes using the BR1 (Belgian Reactor 1) facilities for various research and calibration purposes. The BR1 has proved to be an excellent for calibration and validation of techniques, integral nuclear data validation, activation analysis, characterisation of materials by neutron transmission, and physics experiments. Moreover, the knowledge, built up at BR1 has lead to the best calibration conditions for applying fuzzy logic control (FLC) for nuclear reactor control

  10. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  11. Detailed description of an SSAC at the facility level for on-load refueled power reactor facilities

    International Nuclear Information System (INIS)

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in an on-load refueled power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  12. Decommissioning of the Spent Fuel Storage at the RA Reactor Facility, Serbia

    International Nuclear Information System (INIS)

    Nuclear research reactor RA was constructed in the second half of the 1950s. It was designed in the former Union of Soviet Socialist Republics (USSR), where the main components were also manufactured. The reactor became the largest research nuclear facility in the former Yugoslavia, and was a multipurpose research reactor providing a relatively high neutron flux in the core. It belonged to the second generation of research reactors that gave an important contribution to nuclear technology development in the country. The RA reactor was a tank type reactor using heavy water as a primary coolant and as a moderator. The primary cooling system circulated heavy water to cool the fuel elements in the core and remove heat by upward forced circulation. Its nominal power was 6.5 MW. The facility went critical in December 1959 and was temporarily shut down in August 1984. During this period of operation, the reactor was successfully used for scientific research, but also for commercial purposes. From its first commissioning in 1960, until 1975, the reactor used low enriched uranium fuel (2% of 235U). In 1976, the original fuel was gradually replaced by a high enriched fuel (80% of 235U) that was developed and qualified in the former USSR. After temporary shutdown in 1984, followed by a set of thorough examinations of its systems and equipment, it was decided to reconstruct the reactor systems to enable safe and continuous operation in the future. The reconstruction, with financial help from the IAEA, started in 1986, but owing to international sanctions imposed upon the former Yugoslavia in 1992, the reconstruction work has never been finished. The facility was then left in an extended shutdown regime under passive care and maintenance

  13. Radiological and environmental aspects of fast reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    Availability of energy is an important prerequisite for the socioeconomic development of any country. As the sources of fossil fuels are dwindling fast, India will have to look for nuclear power to secure a stable supply of energy. The Indian nuclear power program aims at large scale utilization of its vast thorium resources. The energy potential of uranium increases by 150 times and that of thorium by three times through the fast breeder reactor route compared to thermal reactors. This long term objective of thorium utilization is sought to be achieved through three stages of development. In the first stage a series of PHWRs will be constructed for power generation which will incidentally generate plutonium. The second stage consists of FBRs with plutonium as the fuel and thorium as the blanket. In the third stage, U-233 will replace plutonium as the fuel for FBRS. It is interesting to compare the radiological and environmental safety aspects of fast reactor fuel cycle involving U-Pu and Th-U

  14. Characteristics and facilities of a 3MW LEU fuelled TRIGA reactor

    International Nuclear Information System (INIS)

    A 3 MW TRIGA reactor fuelled with low enriched uranium having 19.7 % U-235 and 20 wt% Uranium and Zirconium Hydride, has been installed and recently made critical at a research laboratory of the Bangladesh Atomic Energy Commission. This paper describes the basic design, low and high power test results and the facilities of the reactor. The details of the core configuration of the initial criticality with 50 elements and the final core with 100 elements at 3 MW power are discussed. The available experimental facilities are also described briefly. (author)

  15. Severe-fuel-damage experiments in the Canadian in-reactor Blowdown Test Facility

    International Nuclear Information System (INIS)

    The Blowdown Test Facility consists of an instrumented in-reactor irradiation site plus an out-reactor piping system. These are used to irradiate CANDU fuel under conditions representative of a loss-of-coolant accident (LOCA) or LOCA with Loss-Of-Emergency-Core-Cooling (LOECC) in order to study fuel performance, fission-product release from the fuel and the transport of fission products through the piping system. An overview of the facility and the experimental program is given in this paper. (author)

  16. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    International Nuclear Information System (INIS)

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel

  17. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  18. Thermal hydraulic modelling of the Mo and Iridium irradiation facilities of the RA10 reactor

    International Nuclear Information System (INIS)

    The RA-10 reactor is a multipurpose, open pool research reactor. The core consists of a rectangular array of MTR type fuel. The produced thermal power is 30 MW which is extracted by the refrigeration system via an ascendant flow through the core. The core reflector is D2O contained in a watertight tank. The design of the reactor includes a number of out-core facilities which are meant to be used for industrial, medical and research purposes. Among all the facilities, the most important ones are the Molybdenum and Iridium ones which we modeled in this work. During the normal operation of the reactor, the manipulation and the on-line extraction of the irradiation facilities is foreseen. Therefore the study of the head loss during the normal operation as well as during the extraction maneuvers plays a relevant role in the design and safety analysis. In this work a CFD commercial code is use dto perform the calculations needed to guarantee the design requirements.In addition, a full detailed geometric model for both, the Molybdenum and Iridium facilities,is used to perform the required simulations. The obtained results allow to evaluating the thermal-hydraulic performance of the proposed facilities designs. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

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

  1. Predictive maintenance and its use in TRIGA-Pitesti reactor facilities

    International Nuclear Information System (INIS)

    The Pitesti TRIGA reactor is a research and material testing reactor situated on the bottom of an open pool of 300m3, whose steady state nominal power is 14 MW. It is cooled by a primary cooling system which comprises: 4 pumps (2 in operation, 2 in standby) and 3 heat exchangers. The generated heat in the reactor core is removed by a secondary circuit with forced convection towers (provided with 6 ventilators). The reactor was used for complete CANDU fuel testing, structural material (steel, zircaloy) testing and isotope production. The TRIGA Material Test and Research reactor was commissioned at the beginning of 1980. Since that there were made extensive tests on CANDU type fuel and structural materials. It is needed the increase the reliability of equipment's and demanded an improved performance of our facilities. Good maintenance is seen as one of the main keys to improve the performance of TRIGA reactors. For a better operation we are obliged to find and use each up to date methods and strategies. Among these new techniques we could quote the probabilistic assessments, and some of predictive maintenance's techniques. Probabilistic safety and statistical analysis provided useful insights for our reactor operation. During the reactor operation there were unexpected shutdowns, reactor components failures. The data collected were statistically processed in order to obtain a reliability data base. This paper does, indifferently the cause, analysis the failures. The study emphasizes that the most reactor's scrams took place on the first year of work. The scrams number began to lower thereafter and at the end of eighties began to increase again. The greatest number of scrams were caused by the reactor electrical control and instrumentation. An important number of scrams were caused by the irradiation devices. The main conclusion of this study is that the insights are very useful to our operational procedures, to improve the maintenance strategy and the logbook

  2. Safety performance of the fast reactor reprocessing plant and support facilities 1988-1994

    International Nuclear Information System (INIS)

    The Fast Reactor (FR) 'raison d'etre', and the reprocessing and related facilities are briefly described together with the changes in operating philosophy and objectives over the five years mentioned (1988-1994). The safety performance (and difficulties) of a wide spectrum of the facilities - addressing both conventional and radiological hazards - is presented and discussed. The event and incident reporting arrangements are discussed, and some are outlined where they led to significant changes in practice

  3. Development of Facilities and Provision of Radioisotope and Radiopharmaceutical Process based reactor GA Siwabessy

    International Nuclear Information System (INIS)

    The application of radioisotope technology, which is dynamically developing in Indonesia, needs to be compensated by the improvement of capacity and part of the CDRR in preparation and supply of GA Siwabessy-reactor-based radioisotopes and radiopharmaceuticals. On the other hand, the improvement of reactor-based radioisotope production capability will support and push the operational performance and function of the GA Siwabessy reactor facility. Implementation of radioisotope and radiopharmaceutical production capability should be successively joined with the readiness of supporting facility operational function in order to gain safety of the system, process, personnel and environment as well. The aging systems and facilities should be rejuvenated to keep maintenance their operational performance and function. By the reasons of those, some activities have been carried out covering services on operation and maintenance of supporting facilities, services on preparation of GA Siwabessy-reactor-based radioisotope and radiopharmaceutical products, and rejuvenation of chiller system that was gradually performed without disturbing the whole working function of the system. These activities are oriented into improvement of domestic radioisotope application for public welfare especially in the field of health, as well as gaining implementation of whole CDRR's programs safely and pleasantly. The outputs of the activities are merits (operational function, maintenance and repair of the systems and facilities), products (radioisotope and radiopharmaceutical preparations) and rejuvenation of facility. The results of work are quantitatively expressed in the total kinds or amounts of radioisotope and radiopharmaceutical products and the total operational hours of the facilities. Qualitatively, the result of works showed the continuity of readiness of supporting facilities as well as the continuity of services on preparation and supply of radioisotope and radiopharmaceutical

  4. Dual-purpose RTOF diffractometer facility at the ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Some new features of an RTOF diffractometer facility, to be installed at the ET-RR-1 reactor, are considered. The suggested facility will be a dual-purpose instrument. It can be used for high-resolution crystallography utilizing transmission diffraction technique and for strain measurements of technical components using a scattering geometry with a detector at 90 deg. (orig.). (30 refs., 1 fig.)

  5. Safety review and assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    More operational events were occurred at various research reactors in 1995. The NNSA and its regional offices conducted careful investigation and strict regulation. In order to analyze comprehensively the safety situation of inservice research reactors and find same countermeasures the NNSA convened a meeting of the safety regulation on research reactors and a meeting for change experience of the safety regulation on research reactors that were participated in by the operating organizations in 1995. A lot of work has been done in the respects of propagation of regulations on nuclear safety, education of nuclear safety culture, the investigation and treatment of operational events, the reexamine of operation documents, the implementation of rectifying items on nuclear safety, the daily inspection and routine inspection on nuclear safety and the studying on the extending service life of research reactors etc

  6. TRIGA MARK II first research reactor facility in Kingdom of Morocco

    International Nuclear Information System (INIS)

    The research reactor facility is located at Centre d'Etudes Nucleaires de la Maamora(CENM), located approximately 25 kilometers north of the city of Rabat. This facility will enable CNESTEN, as the operating organization, to fulfil its missions for promotion of nuclear technology in Morocco, contribute to the implementation of a national nuclear power program, and assist the state in monitoring nuclear activities for protection of the public and environment. The reactor building include TRIGA Mark II research reactor with an initial power level of 2000kW (t), and equipped for a planned future upgrade to 3,000-kilowatts.The facility is the keystone structure of CENM, and contain in addition to the TRIGA research reactor, extensively equipped laboratories and all associate support systems, structures, and supply facilities with the support of the AIEA, French CEA and LLNL (USA). The CENM with its TRIGA reactor and fully equipped laboratories will give the kingdom of Morocco its first nuclear installation with extensive capabilities. These will include the production of radioisotopes for medical, industrial and environmental uses, metallurgy and chemistry, implementation of nuclear analytical techniques such as neutron activation analysis and non-destructive examination techniques, as well as carrying out basic research programs in solid state and reactor physics. The feedback from the commissioning and the implementation of the safety standards during this phase was very interesting from safety point of view. The TRIGA Mark II research reactor at CENM achieved initial criticality on May 2, 2007 at 13:30 with 71 fuel elements and culminated with the successful completion of the full power endurance testing on 6 September, 2007.

  7. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  8. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  9. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  10. Upgrading of neutron radiography/tomography facility at research reactor

    International Nuclear Information System (INIS)

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  11. SANS facility at the Pitesti 14MW Triga reactor

    International Nuclear Information System (INIS)

    The SANS configuration with mechanical monochromator has significantly increased luminosity while the spatial extension of the instrument is quite reasonable. Tacking account that TRIGA reactor existing at INR Pitesti is a medium flux reactor and that the available dimension for the SANS instrument is severely limited by the dimensions of the room where the instrument has to be installed, this experimental configuration has been chosen as the most suited for the situation existing in our institute. For usual collimation values of about 30 minutes, and for an inclination angle of the monochromator axis of about 2-3 degrees, Dl/l is about 20-30%, i.e. quite reasonable value. The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Qmin = 0.005 A-0-1 while the maximal value is Qmax = 0.5 A-0-1. The relative error is ΔQ/Qmin = 0.5. In the case of our SANS instrument a monochromatic neutron beam with 1.5 A-0 ≤ λ ≤ 5 A-0 is produced by a mechanical velocity selector with helical slots. The distance between sample and detectors plane is (5.2 m). (author)

  12. A complete fuel development facility utilizing a dual core TRIGA reactor system

    International Nuclear Information System (INIS)

    A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 1014 n/cm2-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 1017 n/cm2-sec. The pulse width at half maximum during a

  13. Proposed Fuel Pin Irradiation Facilities for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from spent nuclear fuel. The proposed fuels include metal and oxide mixed actinides (U-Np-Pu-Am-Cm) as well as target concepts with perhaps only Am-Cm. The High Flux Isotope Reactor was built for the purpose of transmuting plutonium to various higher actinides including Am, Cm, and Cf. Since a fast-spectrum irradiation facility does not exist in the United States, HFIR can fulfill a first step in the GNEP mission; that being to establish a near-term capability to irradiate materials in a fast neutron spectrum in addition to efforts to gain access to international facilities through partnering arrangements. Modifications to the HFIR central target region to accomplish this goal are described. A second on-going project for HFIR is to design capsules and installation tools and procedures to irradiate short rods of innovative nuclear fuel types and cladding materials under prototypic LWR operating conditions at an accelerated rate relative to expected reactor performance. This second proposal would be for a facility representative of thermal reactor conditions rather than the GNEP concept. In order to maintain power densities within the fuel at levels normally seen by LWR reactors, an entirely new experiment and test capsule design will be needed than has been available in the past

  14. Education and training at the Rensselaer Polytechnic Institute reactor critical facility

    International Nuclear Information System (INIS)

    The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

  15. Removal of reactor cooling system facilities and others in reconstruction works of JRR-3

    International Nuclear Information System (INIS)

    The home-manufactured No.1 reactor 'JRR-3' in Tokai Research Establishment, Japan Atomic Energy Research Institute, stopped its operation in March, 1983, with the results of cumulative operation time 47,135.5 hours and cumulative power output 419,073 MWh. Since then, the reconstruction works to construct a research reactor with higher performance has been advanced, and the new reactor has attained the initial criticality on March 22, 1990. The removal of the facilities which are not used after the reconstruction was carried out since August, 1985, and in this report, the removal of the parts with relatively high dose equivalent such as reactor cooling system facilities is outlined, except the reactor proper. The range of the removed facilities, the planning of removal works, the progress of removal works and the results are reported. About 110 t of large equipments were preserved as they are, and about 400 t of pipings and others were dismantled, removed, cut and put in containers. (K.I.)

  16. Neutron scattering facilities at the research reactor DR3

    International Nuclear Information System (INIS)

    DR3 is a heavy-water-moderated 10 MW thermal neutron research reactor. The 26 fuel elements contain 2.5-3.5 kg uranium enriched to less than 20% 235U. Neutron beams emerge from four horizontal through-tubes tangential to the reactor core. Two of the horizontal tubes are used for neutron scattering experiments in the field of materials research. The vertical tubes are predominantly used for isotope production and materials testing. The thermal neutron flux is about 3.5x1013 n/cm2/s in the centre of the 7-inch diameter horizontal through-tubes. The thermal neutron flux in equilibrium with the D2O moderator (50 deg. C) has a nearly Maxwellian distribution peaking at 1.1 A. At the maximum flux position in the two horizontal through-tubes used for materials research are installed scatterers designed with a considerably higher scattering power for thermal neutrons than for fast neutrons and gamma-rays. The scatterer is either a 10 mm slab of light water, providing a nearly thermal Maxweellian spectrum at the beam port, or a chamber filled with supercritical hydrogen gas at 16 atmospheres and 38 K, a so-called cold neutron source. The spectrum from a cold source has a considerable flux enhancement in the long wavelength region when compared to the thermal water scatterer. Neutron beams are available for materials research from two thermal and two cold beam ports in the Reactor Hall. One of the cold beams is shared with a 20 meter long cold-neutron guide-tube which provides three beam ports in a separate building, the Neutron House, with could neutrons. Only neutrons that have undergone total reflection from the Ni-coated glass plates in the bent guide-tube arrive at the end of the guide tube in the Neutron House. The angle of total reflection is proportional to the neutron wavelength. Therefore almost no neutrons of wavelength below a certain ''critical'' wavelength are transmitted through the guide-tube, and the experimental equipment installed in the Neutron Houyse

  17. Analyses in support of the Laboratory Microfusion Facility and ICF commercial reactor designs

    International Nuclear Information System (INIS)

    Our work on this contract was divided into two major categories; two thirds of the total effort was in support of the Laboratory Microfusion Facility (LMF), and one third of the effort was in support of Inertial Confinement Fusion (ICF) commercial reactors. This final report includes copies of the formal reports, memoranda, and viewgraph presentations that were completed under this contract

  18. Irradiation facility at the IBR-2 reactor for investigation of material radiation hardness

    Science.gov (United States)

    Bulavin, M.; Cheplakov, A.; Kukhtin, V.; Kulagin, E.; Kulikov, S.; Shabalin, E.; Verkhoglyadov, A.

    2015-01-01

    Description of the irradiation facility and available parameters of the neutron and gamma exposures including the maximal integrated doses are presented in the paper. The research capabilities for radiation hardness tests of materials in high intensity beam of fast neutrons at the IBR-2 reactor of the Joint Institute for Nuclear Research in Dubna (Russia) are outlined.

  19. Irradiation facility at the IBR-2 reactor for investigating material radiation hardness

    Science.gov (United States)

    Bulavin, M. V.; Verkhoglyadov, A. E.; Kulikov, S. A.; Kulagin, E. N.; Kukhtin, V. V.; Cheplakov, A. P.; Shabalin, E. P.

    2015-03-01

    A description of the irradiation facility and available parameters of neutron and gamma exposures, including the maximum integrated doses, are presented in the paper. The research capabilities for radiation hardness tests of materials in a high-intensity beam of fast neutrons at the IBR-2 reactor of the Joint Institute for Nuclear Research in Dubna (Russia) are outlined.

  20. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  1. The Text of the Agreement for the Application of Agency Safeguards to United States Reactor Facilities

    International Nuclear Information System (INIS)

    The text of the Agreement between the Agency and the Government of the United States of America for the application of Agency safeguards to United States reactor facilities, which was signed on 15 June 1964 and entered into force on 1 August 1964, is reproduced in this document for the information of all Members

  2. Neutrons and Gamma-Ray Dose Calculations in Subcritical Reactor Facility Using MCNP

    Directory of Open Access Journals (Sweden)

    Ned Xoubi

    2016-06-01

    Full Text Available In nuclear experimental, training and teaching laboratories such as a subcritical reactor facility, huge measures of external radiation doses could be caused by neutron and gamma radiation. It becomes imperative to place the health and safety of staff and students in the reactor facility under proper scrutiny. The protection of these individuals against ionization radiation is facilitated by expected dose mapping and shielding calculations. A three-dimensional (3D Monte Carlo model was developed to calculate the dose rate from neutrons and gamma, using the ANSI/ANS-6.1.1 and the ICRP-74 flux-to-dose conversion factors. Estimation for the dose was conducted across 39 areas located throughout the reactor hall of the facility and its training platform. It was found that the range of the dose rate magnitude is between 7.50 E−01 μSv/h and 1.96 E−04 μSv/h in normal operation mode. During reactor start-up/shut-down mode, it was observed that a large area of the facility can experience exposure to a significant radiation field. This field ranges from 2.99 E+03 μSv/h to 3.12 E+01 μSv/h. There exists no noticeable disparity between results using the ICRP-74 or ANSI/ANS-6.1.1 flux-to-dose rate conversion factors. It was found that the dose rate due to gamma rays is higher than that of neutrons.

  3. Proposed fuel pin irradiation facilities for the high flux isotope reactor

    International Nuclear Information System (INIS)

    The Global Nuclear Energy Partnership (GNEP) is proposing to develop a sodium-cooled fast-spectrum reactor (SFR) to transmute and consume actinides from spent nuclear fuel. The proposed fuels include metal and oxide forms mixed actinides (U-Np-Pu-Am-Cm) as well as target concepts with perhaps both Am-Cm. The High Flux Isotope Reactor (HFIR) was built for the purpose of transmuting plutonium to various higher actinides including Am, Cm, and Cf Since a fast-spectrum irradiation facility does not exist in the United States, HFIR can fulfill a first step in the GNEP- mission that being to establish a near-term domestic capability to irradiate materials in a fast neutron spectrum. Modifications to the HFIR central target region to accomplish this goal are described. A second ongoing project for HFIR is to design capsules and installation tools and procedures to irradiate short rods of innovative nuclear fuel types and cladding materials under prototypic light water reactor (LWR) operating conditions at an accelerated rate relative to expected reactor performance. This second proposal would be for a facility representative of thermal reactor conditions rather than the GNEP concept. In order to maintain power densities within the fuel at levels normally seen by LWR reactors, an entirely new experiment and test capsule design will be needed. (authors)

  4. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    International Nuclear Information System (INIS)

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  5. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  6. The advanced test reactor national scientific user facility: advancing nuclear technology education

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy designated the Idaho National Laboratory (INL) Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The ATR NSUF provides education programs including a Users Week, internships, faculty student team projects and faculty/staff exchanges. In addition, the ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  7. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  8. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  9. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  10. A comparison of different neutron spectroscopy systems at the reactor facility VENUS

    CERN Document Server

    Vanhavere, F; Chartier, J L; Itie, C; Rosenstock, W; Koeble, T; D'Errico, F

    2002-01-01

    The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed. A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.

  11. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  12. Research on reactor physics using the Japan Materials Testing Reactor Critical Facility (JMTRC)

    International Nuclear Information System (INIS)

    The JMTRC of 100 W was installed for the purpose of carrying out the basic experiment on the nuclear characteristics of reactors and the preceding test related to the operation plan of the Japan material testing reactor (JMTR, 50 MW). After the attainment of the initial criticality in October, 1965, for obtaining the reactor physics characteristics, criticality experiment was begun. The items of the criticality experiment were critical mass, control rod worth, reactor dynamic characteristic parameters, shutdown margin and so on, and these experimental data were effectively utilized for the safety evaluation in the operation of the JMTR. The preceding test using the JMTRC has been carried out for obtaining the nuclear characteristics of samples and the thermal characteristics estimated from those results by simulating the JMTR core. In August, 1983, the degree of fuel enrichment for the JMTRC was reduced to 45 % U-235, and various experiments usig the MEU core were carried out. In this paper, the criticality experiment using the MEU core and the experiment on the characteristics of lithium-containing pellets are reported. (K.I.)

  13. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  14. The University of Missouri Research Reactor facility can melter system

    International Nuclear Information System (INIS)

    At the University of Missouri Research Reactor (MURR), a waste compacting system for reducing the volume of radioactive aluminum cans has been designed, built and put into operation. In MURR's programs of producing radioisotopes and transmutation doping of silicon, a large volume of radioactive aluminum cans is generated. The Can Melter System (CMS) consists of a sorting station, a can masher, an electric furnace and a gas fired furnace. This system reduces the cans and other radioactive metal into barrels of solid metal close to theoretical density. The CMS has been in operation at the MURR now for over two years. Twelve hundred cu ft of cans and other metals have been reduced into 150 cu ft of shipable waste. The construction cost of the CMS was $4950.84 plus 1680 man hours of labor, and the operating cost of the CMS is $18/lb. The radiation exposure to the operator is 8.6 mR/cu ft. The yearly operating savings is $30,000. 20 figs., 10 tabs

  15. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  16. A midsize reactor facility - A regional resource for research and education

    International Nuclear Information System (INIS)

    The mission of the University of Florida Training Reactor (UFTR) is to serve the regional needs of Florida and the Southeast for access to quality reactor usage. Well-advertised capabilities of the facility support diversified usages that include education, training, research, service, and public information programs to address the needs of a broad spectrum of users ranging from high school students and teachers, to university researchers, and even the occasional service user. Despite the midsize power of the facility, the UFTR's status as the only nonpower reactor within 350 miles in one of our largest states means that it is uniquely situated to contribute in these various areas in ways usually reserved for larger facilities. Nine state universities and a well-developed community college system in addition to private schools and a growing complement of progressive high schools assure a broad-based user community. The key to accomplishing mission objectives is to continue diversification and improvement of both the reactor and associated experimental capabilities to meet the needs of this user community

  17. 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

  18. Design and fabrication of sodium test facility for fast breeder reactor

    International Nuclear Information System (INIS)

    The purpose of the promotion policy for energy research and development base construction plan (priority facility) of the Japanese government in FY2009 is 'to construct in Tsuruga City the research and development base for plant operation technology for the practical use of fast breeder reactor where researchers in and out of Japan gather, and to contribute to the development and revitalization of the region as the base with international characteristics.' In conformity to this purpose, the Japan Atomic Energy Agency built 'sodium engineering research facilities' in Tsuruga. This paper describes the design, fabrication, and installation of interior equipment that were carried out by Kawasaki Heavy Industries. 'Sodium engineering research facilities' are the test and research facilities to conduct research and development related to sodium, while reflecting the experiences of operation and maintenance of 'Monju,' which aims at the commercialization of fast reactor. The facilities specialize in the handling technology of sodium to meet the needs in and out of Japan, and were completed in June 2015. The facilities consist of six units including tank-loop test equipment, mini-loop test equipment, sodium purification and supply equipment, etc. For the tank-loop test equipment, a sodium transfer test of about 5.5 tons, and a subsequent comprehensive function test using sodium are scheduled. (A.O.)

  19. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an

  20. TIT reactor laboratory course using JAERI and PNC large experimental facilities

    International Nuclear Information System (INIS)

    This report is presented on a reactor laboratory course for graduate students using large facilities in national laboratories in Japan. A reactor laboratory course is offered every summer since 1990 for all graduate students in the Nuclear Engineering Course in Tokyo Institute of Technology (TIT), where the students can choose one of the experiments prepared at Japan Atomic Energy Research Institute (JAERI), Power Reactor and Nuclear Fuel Development Corporation (PNC) and Research Reactor Institute, Kyoto University (KUR). Both JAERI and PNC belong to Science and Technology Agency (STA). This is the first university curriculum of nuclear engineering using the facilities owned by the STA laboratories. This type of collaboration is promoted in the new Long-Term Program for Research, Development and Utilization of Nuclear Energy adopted by Atomic Energy Commission. Most students taking this course reported that they could learn so much about reactor physics and engineering in this course and the experiment done in large laboratory was a very good experience for them. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

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

  3. USE OF CEMENTITIOUS MATERIALS FOR SRS REACTOR FACILITY IN-SITU DECOMMISSIONING - 11620

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-12-07

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes

  4. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  5. A blueprint for GNEP advanced burner reactor startup fuel fabrication facility

    International Nuclear Information System (INIS)

    Research highlights: → This article discusses use of WG-plutonium as the startup fuel for Advanced Burner Reactor. → The presence of gallium in WG fuel may compromise the fuel integrity. → There is no facility exists to remove gallium from plutonium except at laboratory scale. → This article discusses the processes and issues associated with the gallium removal. → The article provides realistic scenario to all stack-holders involved in designing and operating ABR. - Abstract: The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu)-239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the

  6. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    International Nuclear Information System (INIS)

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  7. Present status of the medical irradiation facility at the Musashi reactor

    International Nuclear Information System (INIS)

    Boron neutron capture therapy (BNCT) of malignant brain tumors has been efficiently performed since March 1977, and the first human case of malignant melanoma was also successfully treated on July 1987 in our reactor (Musashi reactor, TRIGA-II, 100 kW). To obtain both good irradiation field characteristics and a better irradiation facility, some tests and developments have been continued in accordance with the study of medical and biological irradiations. The results of these evaluations and a new approach are presented.6 references

  8. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    International Nuclear Information System (INIS)

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  9. Use of Cementitious Materials for SRS Reactor Facility In-Situ Decommissioning

    International Nuclear Information System (INIS)

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD project requires approximately 250000 cubic yards of cementitious materials to fill the below-grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Funding is being provided under the American Recovery and Reinvestment Act (ARRA). Cementitious materials were designed for the following applications: (A) Below-grade massive voids / rooms: Portland cement-based structural flowable fills for: (A.1) Bulk filling; (A.2) Restricted placement and (A.3) Underwater placement. (B) Special below-grade applications for reduced load bearing capacity needs: (B.1) Cellular portland cement lightweight fill. (C) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels (C.1) Blended calcium aluminate - calcium sulfate based flowable fill; (C.2) Magnesium potassium phosphate flowable fill. (D) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: (D.1) Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured

  10. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  11. Applicability of base-isolation R ampersand D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R ampersand D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R ampersand D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R ampersand D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R ampersand D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs

  12. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  13. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  14. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  15. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  16. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  17. Independent Confirmatory Survey Summary and Results for the Plum Brook Reactor Facility Sandusky OH

    International Nuclear Information System (INIS)

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics (later known as the National Aeronautics and Space Administration or NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities. After successfully completing the objective of landing humans on the Moon and returning them safely to Earth, NASA was faced with budget reductions from Congress in 1973. These budgetary constraints caused NASA to cease operations at several research facilities across the country, including those at Plum Brook Station. The major test facilities at Plum Brook were maintained in a standby mode, capable of being reactivated for future use. The Plum Brook Reactor Facility (PBRF) was shut down January 5, 1973 and all of the nuclear fuel was eventually removed and shipped off site to a U.S. Department of Energy facility in Idaho for disposal or reuse. Decommissioning activities are currently underway at the PBRF (NASA 1999). The objectives of the confirmatory survey activities were to provide independent contractor field data reviews and to generate independent radiological data for use by the Nuclear Regulatory Commission (NRC) in evaluating the adequacy and accuracy of the licensee's procedures and final status survey (FSS) results

  18. Characterization and adjustment of the neutron radiography facility of the RP-10 nuclear reactor

    International Nuclear Information System (INIS)

    The main aim of this work was to characterize and adjust the neutron radiography facility of the RP-10 nuclear reactor, and therefore be able to offer with this technique services to the industry and research centers in general. This technique will be complemented with others such as x-rays and gamma radiography. First, the shielding capacity of the facility was analyzed, proving that it complies with the radiological safety requirements established by the radiological safety code. Then gamma filtration tests were conducted in order to implement the direct method for image formation, optical density curves were built according to the thickness of the gamma filter, the type of film and the type of irradiation. Also, the indirect method for image formation was implemented for two types of converters: indium and dysprosium. Growth curves for optical density were also made according to contact time between converter-film, for different types of films. The resolution of the facility was also analyzed using two methods: Klasens (1946) and Harms (1986). Harms method came closer to the resolution of the human eye when compared to the Klasens method. Finally, the application fields of neutron radiography are presented, including those conducted at the neutron radiography facility of the RP-10 nuclear reactor. With this work, the RP-10 neutron radiography facility is ready to offer inspection and research services

  19. Preliminary Neutronics Design for Fast Neutron Irradiation Facility in the KIJANG Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hongchul; Seo, Chul Gyo; Chae, Hee Taek; Kim, Myongseop [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Current design of FNI facility in the KJRR was introduced in this paper. The performance evaluations for the FNI facility were carried out and the results meet the requirements. There is not a constraint how far gamma flux should be lowered, and the further study on this subject should be researched. The HANARO research reactor has been successfully operated in Korea. Following the experience with HANARO, KIJANG research reactor (KJRR) is now being designed by Korea Atomic Energy Research Institute (KAERI), dedicated to increasing the national radio-isotopes supply capacity including the self sufficiency of Mo-99. And the KJRR is also expected to have the capability to provide the neutron irradiation service for power semiconductor production in a large scale. This service includes not only neutron transmutation doping (NTD) facility for ingot irradiation, but also fast neutron irradiation (FNI) facility for wafer irradiation. Fast neutron irradiation for wafer is a promising technology for efficiency gain and life extension of the power semiconductor. In this work, a FNI facility has been preliminarily designed to satisfy its requirements, and the necessary calculations are carried out through Monte Carlo simulations using MCNP5 code.

  20. Large sample NAA facility at GRR-1 research reactor: Design and applications

    International Nuclear Information System (INIS)

    Full text: A Large Sample Neutron Activation Analysis (LSNAA) facility is under development at GRR-1 research reactor, NCSR 'Demokritos'. The LSNAA facility design incorporates sample irradiation in the reactor's graphite thermal neutron column and subsequent measurement of the activity induced at a gamma spectroscopy system with gamma ray transmission measurement options included. Monte Carlo neutron and photon transport code MCNP-4C was used to model the facility. Appropriate correction factors accounting for neutron field perturbation during sample irradiation, high purity germanium detector efficiency for the volume source and gamma ray self-absorption within the sample itself were derived. The results of the computations were experimentally verified by activation foil measurements for a set of known materials and a range of sample sizes extending up to 10 litters. Moreover, the special issue of large sample analysis of non-homogeneous samples is examined and the limits of the technique are discussed. The LSNAA facility will be used to perform multi-element, non-destructive, contamination free analysis of large volume samples with high sensitivity and excellent sampling. End-users of the facility include archaeological, environmental, biomedical research laboratories and the industry. Preliminary results of large sample analysis are presented and the accuracy of the technique is validated against conventional instrumental neutron activation analysis of the same materials. (author)

  1. Predictive maintenance and its use in TRIGA-Pitesti reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Preda, M.; Barbalata, E.; Sabau, C. (Institute for Nuclear Research, Pitesti (Romania))

    1999-12-15

    The Pitesti TRIGA reactor is a research and material testing reactor situated on the bottom of an open pool of 300m3, whose steady state nominal power is 14 MW. It is cooled by a primary cooling system which comprises: 4 pumps (2 in operation, 2 in standby) and 3 heat exchangers. The generated heat in the reactor core is removed by a secondary circuit with forced convection towers (provided with 6 ventilators). The reactor was used for complete CANDU fuel testing, structural material (steel, zircaloy) testing and isotope production. The TRIGA Material Test and Research reactor was commissioned at the beginning of 1980. Since that there were made extensive tests on CANDU type fuel and structural materials. It is needed the increase the reliability of equipment's and demanded an improved performance of our facilities. Good maintenance is seen as one of the main keys to improve the performance of TRIGA reactors. For a better operation we are obliged to find and use each up to date methods and strategies. Among these new techniques we could quote the probabilistic assessments, and some of predictive maintenance's techniques. Probabilistic safety and statistical analysis provided useful insights for our reactor operation. During the reactor operation there were unexpected shutdowns, reactor components failures. The data collected were statistically processed in order to obtain a reliability data base. This paper does, indifferently the cause, analysis the failures. The study emphasizes that the most reactor's scrams took place on the first year of work. The scrams number began to lower thereafter and at the end of eighties began to increase again. The greatest number of scrams were caused by the reactor electrical control and instrumentation. An important number of scrams were caused by the irradiation devices. The main conclusion of this study is that the insights are very useful to our operational procedures, to improve the maintenance strategy and the

  2. Facility at CIRUS reactor for thermal neutron induced prompt γ-ray spectroscopic studies

    Science.gov (United States)

    Biswas, D. C.; Danu, L. S.; Mukhopadhyay, S.; Kinage, L. A.; Prashanth, P. N.; Goswami, A.; Sahu, A. K.; Shaikh, A. M.; Chatterjee, A.; Choudhury, R. K.; Kailas, S.

    2013-03-01

    A facility for prompt γ-ray spectroscopic studies using thermal neutrons from a radial beam line of Canada India Research Utility Services (CIRUS) reactor, Bhabha Atomic Research Centre (BARC), has been developed. To carry out on-line spectroscopy experiments, two clover germanium detectors were used for the measurement of prompt γ rays. For the first time, the prompt γ-γ coincidence technique has been used to study the thermal neutron induced fission fragment spectroscopy (FFS) in 235U(nth, f). Using this facility, experiments have also been carried out for on-line γ-ray spectroscopic studies in 113Cd(nth, γ) reaction.

  3. Detailed description of an SSAC at the facility level for research reactors

    International Nuclear Information System (INIS)

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in a research reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  4. Design of neutron radiography facility in pool for the reactor RA-10

    International Nuclear Information System (INIS)

    RA-10 project consists in the design and construction of a multipurpose reactor for multiple applications, including radioisotopes production, material testing and an in pool facility for neutron imaging. Neutron imaging is a powerful tool for studies of materials and offer several advantages among other attenuation-based techniques. In this study mechanical and neutronic requirements for the RA-10 in pool neutron imaging facility are described. The MCNP neutronic model and the mechanical design satisfying these requirements in a first engineering stage are described. (author)

  5. The small-angle neutron scattering facility at the SAFARI-1 reactor

    International Nuclear Information System (INIS)

    A long-wavelength neutron scattering facility at the SAFARI-1 reactor is described. Neutrons of wavelength between 5 and 15 A can be selected. Features of the facility are the use of microwave guides as neutron conductors, flexible guide-pipe configuration and automatic sequential sample changing. Examples are given of measurements on radiation-induced voids in copper, aluminium, Al-0.4%Si and Al-0.1%In after neutron irradiation and magnetic scattering in US and in 80US-10UC-10UC2. (Auth.)

  6. MVP Based Calculation of Reactivity Loss Due to Gemstone Irradiation Facility of Thai Research Reactor

    International Nuclear Information System (INIS)

    Full text: The calculation of initial core criticality of Thai Research Reactor-1/Modification 1 was performed by the continuous energy Monte Carlo Code MVP and the material cross-sections from JENDL-3.3 continuous-energy library. After that gemstone irradiation facility model were extended for calculation on the magnitude of the reactivity loss. The results showed that total reactivity worth of the control system was 10.83. The reactivity effects associated with the insertion of gemstone irradiation facility was about -0.43% δk/k

  7. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL) - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR- 2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  8. Neutron Beam Characterization for Neutron Radiography Facility at the Thai Research Reactor TRR-1/M1

    International Nuclear Information System (INIS)

    The aim of this research is to characterize the present status of neutron beam coming out from the reactor core of Thai Research Reactor TRR-1/M1 through neutron radiography facility. In this study, the neutron beam profiles at different positions along the beam exit were recorded using digital imaging devices. In addition, thin foil activation technique, with and without cadmium cover, was employed to determine thermal neutron flux and Cd ratio. An acrylic step wedge was exposed to neutron at different time. In parallel to image construction, neutron detection was carried out using a BF3 gas-filled detector. Then, the image intensities at particular thicknesses were normalized by neutron counts from the BF3 detector to determine relative neutron intensity. The obtained information of neutron beam characterization will be useful not only for monitoring the present status of neutron radiography facility but also for determining the optimum exposure conditions for particular samples in the future.

  9. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  10. Research reactor and fuel development/production facility decommissioning technology and experience

    International Nuclear Information System (INIS)

    This paper discusses the technology and experience gained in a series of reactor and fuels development facility decommissioning programs carried out by Babcock and Wilcox (B and W) at its US Nuclear Regulatory Commission (NRC)-licensed sites in Lynchburg, Virginia. Areas of generic application to future projects are particularly emphasized. The projects included one test and one research reactor, four low-power critical experiment facilities, and two buildings that housed plutonium/uranium fuels development laboratories. These projects were comprehensive; they included developing the decommissioning and quality assurance plans, interfacing with the NRC, performing the actual decontamination/dismantling work, performing predecontamination and final radiological surveys, and volume reducing, packaging, certifying, classifying and shipping the radioactive waste for disposal

  11. Student Training Course Using the Experimental Fast Reactor JOYO and Related Facilities

    International Nuclear Information System (INIS)

    University level training courses have been initiated and implemented using the Experimental Fast Reactor Joyo and related facilities of the Japan Atomic Energy Agency (JAEA). These courses offer nuclear facility on-site education and experience in conjunction with a highly experienced engineering staff. University Nuclear Engineering Department faculty members have strongly supported and collaborated in the development of this program. The program covers reactor core physics analysis plus experiments using full-scope training simulator and performing neutron dosimetry, isotopic analysis of noble gases, chemical analysis of sodium, etc. This program is also anticipated to promote the human resource development in the younger generation for the nuclear industry, and to strengthen the relation between JAEA and University research programs. (author)

  12. Sodium natural convection testing in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility

    International Nuclear Information System (INIS)

    A comparison is made between experimental data and analytical results for a single-phase natural convection test in an experimental sodium loop. The test was conducted in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility, an engineering-scale high temperature sodium loop at the Oak Ridge National Laboratory (ORNL), used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies at normal and off-normal operating conditions. Electrical heating in the 19-pin assembly during the test was typical of decay heat levels. The test chosen for analysis in this paper was one of seven natural convection runs conducted in the facility. In this test the bypass line was open to simulate a parallel heated assembly and the test was begun with a pump coastdown from a small initial forced flow

  13. The experience from the construction of BNCT facility at the LVR-15 reactor

    International Nuclear Information System (INIS)

    The BNCT project at LVR-15 reactor of NRI for treatment of human brain gliomas is before start of clinical trials. A survey of present conditions is included, the attention is devoted to BNCT facility with epithermal neutron beam first of all. The different materials for filter composition were studied, the calculational methods have been used for the determination of neutron and gamma rays in the reactor geometry. Some configurations were experimentally verified. The effort for improvement of epithermal neutron beam parameters in configuration 1998 was concentrated to block of filters remodelling, improvement of collimator-shutter geometry, the choice of optimal reactor core edge configuration. Awaited results from experiment in June 1999 are described. (author)

  14. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  15. Neutron Radiography Facility at IBR-2 High Flux Pulsed Reactor: First Results

    Science.gov (United States)

    Kozlenko, D. P.; Kichanov, S. E.; Lukin, E. V.; Rutkauskas, A. V.; Bokuchava, G. D.; Savenko, B. N.; Pakhnevich, A. V.; Rozanov, A. Yu.

    A neutron radiography and tomography facilityhave been developed recently at the IBR-2 high flux pulsed reactor. The facility is operated with the CCD-camera based detector having maximal field of view of 20x20 cm, and the L/D ratio can be varied in the range 200 - 2000. The first results of the radiography and tomography experiments with industrial materials and products, paleontological and geophysical objects, meteorites, are presented.

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

    International Nuclear Information System (INIS)

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

  17. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    International Nuclear Information System (INIS)

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established

  18. Descriptions of selected accidents that have occurred at nuclear reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bertini, H.W.

    1980-04-01

    This report was prepared at the request of the President's Commission on the Accident at Three Mile Island to provide the members of the Commission with some insight into the nature and significance of accidents that have occurred at nuclear reactor facilities in the past. Toward that end, this report presents a brief description of 44 accidents which have occurred throughout the world and which meet at least one of the severity criteria that were established.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

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

  20. A very simple facility for reactor irradiations from 70 - 4000C

    International Nuclear Information System (INIS)

    An irradiation loop was installed at the Research Reactor of 'Demokritos'. This facility allows the irradiation with fast neutrons from 70 to 4000C. A description is given of the loop. As an example of its use, a figure is shown of the negative resistivity curve of Cu + 50 at%Ni as a function of neutron dose at an irradiation temperature of 730C. (U.K.)

  1. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to June 1978

  2. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  3. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    International Nuclear Information System (INIS)

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection

  4. Challenges in the design of Waste Management Plant (WMP) for Fast Reactor Fuel Cycle Facility (FRFCF)

    International Nuclear Information System (INIS)

    A 500 MW, Prototype Fast Breeder Reactor (PFBR) is being constructed by DAE at Kalpakkam, Tamilnadu. Fast Reactor Fuel Cycle Facility (FRFCF) is also being planned as an integrated facility in close proximity to PFBR to cater to the need of closing the fast reactor fuel cycle. The design of waste management plant of FRFCF is one of the important and challenging tasks in FRFCF. The high burn up, short cooling period and high fissile content of spent fuel input from fast Reactors to FRFCF are the main reasons for the difficulties / challenges posed in the management of wastes from FRFCF. Separation of actinides and Platinum Group Metals (PGM) from high level liquid wastes, higher waste oxide loading in vitrified waste product from high level liquid waste management, development of better sorbent and chemical treatment method for achieving higher decontamination factors in the treatment of low and intermediate level liquid wastes and separation of plutonium from spent / de-graded solvent waste are some of the challenges which are addressed in this paper. Solid waste management is also associated with major challenges because of high volume of alpha contaminated wastes. Proper segregation, assaying, conditioning and treatment for volume reduction are planned. Management of significant amount of solid wastes generated as fuel hardware components (alpha and non-alpha) through proper storage and retrieval for future conditioning / disposal are also addressed in this paper. This paper describes the challenges faced during the design and solutions provided for each system. (author)

  5. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  6. Experimental thermal hydraulic facility for simulating LOCA behaviour of pressurised heavy water power reactor

    International Nuclear Information System (INIS)

    Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)

  7. Regulatory Management of Research Reactor Spent Fuel Facilities in Australia- Managing a Cropping Incident

    International Nuclear Information System (INIS)

    The Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) is responsible for the regulation of nuclear installations in Australia under the Australian Radiation Protection and Nuclear Safety Agency Act 1998 (the Act) and the Australian Radiation Protection and Nuclear Safety Regulations 1999 (the Regulations). Nuclear Installations covered by the Act include research reactors, radioisotope production facilities, waste management facilities and fuel management facilities. All of Australia's existing nuclear installations are under the control of the Australian Nuclear Science and Technology Organisation (ANSTO). There are two shut down research reactors that have been principally responsible for Australia's volume of spent fuel. ANSTO operated 10 MW HIFAR for about fifty years. ANSTO also operated an Argonaut type 100 kW reactor (MOATA) for about 34 years with HEU fuel. In the case of the HIFAR reactor the spent fuel arising from operation included both HEU and LEU fuel assemblies. The volume of LEU was much smaller since LEU fuel was only used in HIFAR reactor from 2004 till the time of its final shutdown in early 2007. ANSTO's new 20 MW OPAL research reactor utilises LEU, hence, future volumes of LEU will be greater. The major function of the ANSTO fuel management facility is passive, that is the safe storage of new and spent fuel in engineered purpose-built facilities. The Fuel Management facilities (known as Fuel Operations) principally comprise: the Spent Fuel Wet Stores; the Active Handling Pond where loading of casks with spent fuel takes place for off-site transport, High Activity Handling Cells used for inspecting spent fuel elements, or the handling of other radioactive items; and, the Nuclear Materials Vault and Store used for storing new HIFAR fuel and other fissile material. All of these facilities are located at the Lucas Heights Science and Technology Centre (near Sydney). This works describes the regulatory approach in managing

  8. Requirements for a helium-cooled blanket heat removal system development facility for fusion reactor research

    International Nuclear Information System (INIS)

    Existing and potential design problems associated with the helium-cooled blanket assemblies of experimental, demonstration and hybrid reactor designs considered in the Magnetic Fusion Energy (MFE) Program were assessed. It was observed that a balanced program of design, analysis and experimentation would be required to develop, verify and qualify these designs and those of related hardware and equipment. To respond to the potential experimental requirements of the first-generation reactors (the EPRs and possibly the hybrid concept), the need for a helium test facility was identified. It was determined that this facility should have the capacity for recirculating 100,000 kg/hr of helium at 70 atm and 6000C and should have 3 MW of electrical power available for simulating neutron heating. No radioactive material or processes should be used to facilitate ''hands-on'' experimentation and development. The general types of testing anticipated in this facility would include: (1) thermal and coolant flow performance of the blanket and other components in the primary cooling circuit; (2) structural adequacy of the blanket and first wall including vibration considerations; (3) capability for accommodating safety/off-normal conditions. Existing facilities worldwide were surveyed. It was determined that a number of facilities exist in foreign nations for performing the anticipated experiments. However, no large helium gas flow loop exists within the USA. Consequently, it is recommended that a helium thermal-hydraulic blanket test facility be planned and build on a schedule that will meet the unique design development and verification needs of the fusion program. This report provides the rationale and preliminary scoping of the operational characteristics and requirements for such a facility

  9. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    International Nuclear Information System (INIS)

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  10. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  11. Teaching and training at RA-6 reactor and their contribution to the research facilities development

    International Nuclear Information System (INIS)

    Full text: The RA-6 is a pool type 500 KW reactor, with U 90% enriched fuel. It is located at Bariloche Atomic Centre (Argentina) and it was designed for teaching, training and research purposes. It has been strongly involved in the Nuclear Engineer career of the Balseiro Institute through grade and post grade courses and master and PHD thesis during the last 20 years in topics like: reactor physics, radiation protection, activation analysis, fuel cycle, radiation dosimetry, BNCT, neutron radiography, beam physics and medical isotopes production. It has also provided training for the exploration teams of the NUR (Algeria) and MPRR (Egypt) reactors commissioned in the last decade and for the personnel involved in the development and operation of its facilities. Operators of the Argentinean Nuclear Power Plant Atucha I and Embalse received specific training in reactor physics and more than 10 world wide professionals were trained, supported by IAEA fellowships. The irradiation facilities initially included in the reactor design were improved focusing in specific interests with an appropriate personnel management, and student's co-operation that allowed achieving different grades of development: The Neutronic Activation Analysis Laboratory (LAAN) has been widely used not only for standard applications but also to water contamination and forensic studies. The Boron Neutron Capture Therapy (BNCT) facility has been completed, with an hyperthermal beam which will be soon used for the treatment of human malignant melanomas; but it has been already been used for 'in vivo' and 'in vitro' experiments in hamsters and mousses which contributed to the development of new applications of BNCT. Beam characterization was included in the Dosimetry Exchange project involving the intercomparison of results with other existing facilities in the world. The Prompt Gamma Neutron Activation Analysis (PGNAA) is being upgraded in the frame of the IAEA project 'New Applications on PGNAA'. It

  12. The Budapest research reactor as an advanced research facility for the early 21st century

    International Nuclear Information System (INIS)

    The Budapest Research Reactor, Hungary's first nuclear facility was originally put into operation in 1959. The reactor serves for: basic and applied research, technological and commercial applications, education and training. The main goal of the reactor is to serve neutron research. This unique research possibility is used by a broad user community of Europe. Eight instruments for neutron scattering, radiography and activation analyses are already used, others (e.g. time of flight spectrometer, neutron reflectometer) are being installed. The majority of these instruments will get a much improved utilization when the cold neutron source is put into operation. In 1999 the Budapest Research Reactor was operated for 3129 full power hours in 14 periods. The normal operation period took 234 hours (starting Monday noon and finishing Thursday morning). The entire production for the year 1999 was 1302 MW days. This is a slightly reduced value, due to the installation of the cold neutron source. For the year 2000 a somewhat longer operation is foreseen (near to 4000 hours), as the cold neutron source will be operational. The operation of the reactor is foreseen at least up to the end of the first decade of the 21st century. (author)

  13. Development of sodium facilities for NSRR fast breeder reactor fuel tests. 2. Sodium capsule

    International Nuclear Information System (INIS)

    In order to commercialize fast reactors, which are expected to be long-term transmutes of plutonium and long half life radioactive wastes (such as americium) from light water reactors, safety research under accident conditions and establishment of the safety guidelines are essential. Sodium facilities, such as, (1) Purification/charging loop and test loop, and (2) Proto-type Sodium capsule, were developed and fabricated in order to pulse irradiate fast breeder reactor fuels in the Nuclear Safety Research Reactor (NSRR) of JAERI for investigation on fuel behavior under transient over-power conditions. This report presents the purpose, outlines, specifications, capabilities and operation results of the proto-type sodium capsule. Two kinds of capsule, i.e., the stagnant sodium capsule and the sodium loop, were designed to pulse irradiate Fast Reactor (FR) fuels in the NSRR under sodium cooling conditions with and without flow, respectively. Because the capsules have to safely contain chemically active sodium at high temperature and stand the pressure pulses by the sodium hummer which might be generated at fuel failure, the development of the capsule is essential for realizing the research. Thus, proto-type sodium loop, which consisted of doubly sealed container, sodium pump and flow meter, was developed. In addition, two type of flange structure for the stagnant capsule and loop was leak tested at high pressures, in order to confirm its sealing capability at room and high temperature conditions. (author)

  14. Risk analysis of the aqueous fast reactor fuel cycle facility in the conceptual design stage

    International Nuclear Information System (INIS)

    This paper describes the radioactive release risk of the advanced aqueous reprocessing and fabrication facility for the fast reactor fuel cycle. Because this advanced facility is still in the conceptual design stage, the risk analysis aims at grasping the entire risk comprehensively as simple as possible. As a potential hazard, it was shown that the main process in the reprocessing and fuel fabrication facilities involved only an order of 10-3 of radioactivity in the single reactor core of large scale. Abnormal phenomena related to radioactive solution that can cause radioactive release from the facility to the environmental atmosphere in a large quantity were identified as follows: in-vessel boiling caused by loss of cooling system, a leak and fire of inflammable organic solvent in a cell, in-vessel boiling due to criticality accident, an explosion. Simplified estimation about the quantitative risk of radioactive release showed that in-vessel boiling due to loss of cooling system had the largest contribution to the non-volatile radioactive substance release in a large quantity and that criticality accidents initiated from incomplete extraction stripping of Pu nuclides were dominant in the release risk of radioactive iodine and noble gas with a short-half-life. (author)

  15. Startup of the Fission Converter Epithermal Neutron Irradiation Facility at the MIT Reactor

    International Nuclear Information System (INIS)

    A new epithermal neutron irradiation facility, based on a fission converter assembly placed in the thermal column outside the reactor core, has been put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). This facility was constructed to provide a high-intensity, forward-directed beam for use in neutron capture therapy with an epithermal flux of [approximately equal to]1010 n/cm2.s at the medical room entrance with negligible fast neutron and gamma-ray contamination. The fission converter assembly consists of 10 or 11 MITR fuel elements placed in an aluminum tank and cooled with D2O. Thermal-hydraulic criteria were established based on heat deposition calculations. Various startup tests were performed to verify expected neutronic and thermal-hydraulic behavior. Flow testing showed an almost flat flow distribution across the fuel elements with <5% bypass flow. The total reactivity change caused by operation of the facility was measured at 0.014 ± 0.002% δK/K. Thermal power produced by the facility was measured to be 83.1 ± 4.2 kW. All of these test results satisfied the thermal-hydraulic safety criteria. In addition, radiation shielding design measurements were made that verified design calculations for the neutronic performance

  16. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  17. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11

  18. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  19. Procurement of tritium for fusion reactor. A design study of facility for production of fusion fuel tritium

    International Nuclear Information System (INIS)

    Tritium, a developmental fuel for use in fusion reactors, has been produced in fission research reactors in Japan by extraction from neutron-irradiated 6Li-targets. This paper describes the preliminary design of a large-scale production facility capable of producing 500 g of tritium annually. The present status of tritium production technology in Japan is also discussed. (author)

  20. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Science.gov (United States)

    2012-02-13

    ... COMMISSION Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108... Chemical Company (the licensee) to operate the Dow Chemical TRIGA Research Reactor (DTRR) at a maximum... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process...

  1. Reactor facility for district heating with atmospheric pressure in the primary circuit (RUTA-70)

    International Nuclear Information System (INIS)

    The RUTA-70 is a Russian abbreviation for the reactor facility of 70 MW(th) for district heating with atmospheric pressure in the primary circuit. Single-purpose nuclear heating plants (NHP) are regarded as the possible sources of heat supply for district heating in Russia. A challenging design in this area is the reactor facility RUTA designed specifically as a heat source for district heating systems. A reliable and simple design of the RUTA is ensured, first of all, by the lack of excess pressure in the primary coolant (in the reactor pool). NHPs with such reactors are characterized by inherent safety features and could be located in the immediate vicinity of the heat users. NHPs with the RUTA reactors are suitable for heat supply to urban areas with a p population in the range of 10 to 100 thousand people. In 1990, the Research and Development Institute of Power Engineering of the Rosatom of the Russian Federation (NIKIET) with the participation of IPPE (Obninsk, Russia) has developed the conceptual design of the RUTA plant with a 20 MW(th) output. In 1992, based on this design, the Russian organizations NIKIET, IPPE, VNIPIET and the Mining Institute of the Kola Research Centre of the Russian Academy of Sciences (MI KRC RAS) prepared a feasibility report called 'Designing of the underground NHP with the RUTA reactor for district heating of Apatity, Murmansk region', where it was suggested to use two reactors of 20 MW(th) each for the identified purpose. When discussing these proposals with the town council of Apatity and with the authorities of the Murmansk region, it was considered feasible to develop the design of the RUTA NHP for Apatity and several smaller towns with the unit power output of the reactor increased to 50-60 MW(th). The feasibility report, 'Underground NHP with the RUTA reactors of 4 x 55 MW(th) output for district heating in Apatity, Murmansk region', was prepared in 1994 by joint efforts of specialists from the NIKIET, IPPE, VNIPIET, MI KRC

  2. HFIR [High-Flux Isotope Reactor] irradiation facilities improvements: Completion of the HIFI [High Irradiation Facilities Improvements] project

    International Nuclear Information System (INIS)

    The HFIR Irradiation Facilities Improvements (HIFI) Project has now been completed. In August 1986, Phase I of the project was completed, providing the capability to perform instrumented irradiation experiments in the target region of the HFIR. In June 1987, Phase II of the project was completed with the assembly in the reactor mockup of all the components necessary to operate up to eight 46-mm-diam instrumented experiments in the removable beryllium region of the HFIR. In conjuntion with the installation of Phase I components, the first instrumented target capsule was installed to determine more accurately the probable temperature in the uninstrumented target capsules previously irradiated as part of the Japan/US fusion materials program. Data from this experiment indicate close agreement with expected temperatures in all positions except those at the extreme ends of the capsule. These data provide a more accurate axial gamma heating rate profile that will allow for better design of future HFIR target irradiation capsules

  3. Modular head assembly and method of retrofitting existing nuclear reactor facilities

    International Nuclear Information System (INIS)

    A method is described of retrofitting existing nuclear reactor facilities so as to form a modular closure head assembly for a nuclear reactor pressure vessel, where the existing nuclear reactor facilities comprise control rod drive mechanism cooling systems which include vertically extending elbow air ducts inter-connecting vertically spaced upper and lower air manifolds. The elbow air ducts extend radially beyond the peripheral envelope of the closure head, comprising the steps of: removing the upper air manifold; removing the vertically extending elbow air ducts; capping the air ports of the lower air manifold which ports were previously fluidically connecting the lower air manifold to the vertically extending elbow air ducts; disposing vertically upwardly extending air exhaust ducts above the lower air manifold in such an manner that the air exhaust ducts are disposed within the peripheral envelope of the closure head; fluidically connecting exhaust fans to the upper regions of the air exhaust ducts; fluidically connecting the lower regions of the air exhaust ducts the lower air manifold; permanently securing lift rods to the closure head at positions disposed radially outwardly of the lower air manifold; attaching a seismic support platform to the lift rods; proving fluidic passage of the vertically extending air exhaust ducts through the seismic support platform; attaching a missile shield plate to the lift rods; and proving fluidic passage of the vertically extending air exhaust ducts through the missile shield plate

  4. Applied research and service activities at the University of Missouri Research Reactor Facility (MURR)

    International Nuclear Information System (INIS)

    The University Of Missouri operates MURR to provide an intense source of neutron and gamma radiation for research and applications by experimenters from its four campuses and by experimenters from other universities, government and industry. The 10 MW reactor, which has been operating an average of 155 hours per week for the past eight years, produces thermal neutron fluxes up to 6-7x1014 n/cm2-s in the central flux trap and beamport source fluxes of up to 1.2x1014 n/cm2-s. The mission of the reactor facility, to promote research, education and service, is the same as the overall mission of the university and therefore, applied research and service supported by industrial firms have been welcomed. The university recognized after a few years of reactor operation that in order to build utilization, it would be necessary to develop in-house research programs including people, equipment and activity so that potential users could more easily and quickly obtain the results needed. Nine research areas have been developed to create a broadly based program to support the level of activity needed to justify the cost of operating the facility. Applied research and service generate financial support for about one-half of the annual budget. The applied and service programs provide strong motivation for university/industry association in addition to the income generated. (author)

  5. Alteration of reactor installation (alteration of JMTR reactor facilities and radioactive waste treatment facilities) in the Oarai Research Establishment, Japan Atomic Energy Research Institute (report)

    International Nuclear Information System (INIS)

    The Nuclear Safety Commission presented to the Prime Minister the report on the above alteration after the prudent deliberation, about which the inquiry was made on May 2, 1988. It was recognized as the result of examination that the technical capability of the applicant is appropriate. It was judged as the result of examination that the safety after this alteration of reactor installation can be ensured. The main matters of investigation were as follows. In order to increase the capacity for preserving 200 L drums in which radioactive wastes were enclosed and solidified, a place for accumulating and preserving solids is to be newly installed. It is a one-story reinforced concrete building, and the capacity for preservation, the spatial dose rate around the site, the aseismatic design and so on were investigated. In order to increase the capacity for storing spent ion exchange resin discharged from the primary cooling facility and the irradiation facility of the JMTR, a storage tank is to be installed in the third drain system. The capacity, the countermeasures to leak, the surface dose rate and others were examined. (K.I.)

  6. Development of the irradiation facility for simulating condition of light-water reactor

    International Nuclear Information System (INIS)

    JMTR is an irradiation test reactor of light-water moderated and light-water cooling type with Japan's largest output. It ceased operation once in 2006, due to aging after passage of 38 years from the initial criticality. Thereafter, in order to meet a wide range of utilization needs, the reactor facilities have been renovated for the purpose of re-running since 2007. In addition, aiming at the long-life measures of current light-water reactor, the construction of research basis for problem solving for improving science and technology, and the formation of the core of international research and development, the development of cutting-edge irradiation equipment in parallel with the refurbishment of JMTR are carried out. Of them, this paper takes up the development of irradiation equipment simulating the real environment of right-water reactor, and reports the preparation state of the following items: (1) design of the equipment, (2) fabrication and installation, (3) environmental preparation for installation site, and (4) fabrication of ancillary equipment such as stop flange, connection unit, and capsule. (A.O.)

  7. Planning and management for the decommissioning of research reactors and other small nuclear facilities

    International Nuclear Information System (INIS)

    Many research reactors and other small nuclear facilities throughout the world date from the original nuclear research programmes in the Member States. Consequently, a large number of these plants have either been retired from service or will soon reach the end of their useful lives and are likely to become significant decommissioning tasks for those Members States. In recognition of this situation and in response to considerable interest shown by Member States, the IAEA has produced this document on planning and management for the decommissioning of research reactors and other small nuclear facilities. While not directed specifically at large nuclear installations, it is likely that much of the information presented will also be of interest to those involved in the decommissioning of such facilities. Current views, information and experience on the planning and management of decommissioning projects in Member States were collected and assessed during a Technical Committee Meeting held by the IAEA in Vienna from 29 July to 2 August 1991. It was attended by 22 participants from 14 Member States and one international organization. 28 refs, 2 figs, 3 tabs

  8. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 1010 n/cm2·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10-11 cGy·cm2/nth) and fast neutron (0.58 x 10-11 cGy·cm2/nth) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  9. Considerations on neutron/environment characteristics in inertial fusion ignition (NIF) and reactor (KOYO) facilities

    International Nuclear Information System (INIS)

    This paper reviews reported calculations on the assessment of some environmental questions related to neutron activation in ignition (national ignition facility, NIF) and reactor (KOYO project) inertial fusion facilities. NIF results on the availability of the target chamber after different gain shots, inventories of activated gases in the bay area, and comments on radiological impacts from normal operations have shown the good environmental qualifications of this facility. The laser fusion conceptual reactor KOYO, developed by the ILE Osaka, uses liquid LiPb film protection flowing through ceramic SiC-base porous tubes in two different packing fraction blankets. It is shown the availability of a large fraction of the SiC with impurities to be considered as class C shallow land burial (SLB); however, the complete solution for SLB is the use of porous woven graphite (C) fabric tubes. An important effect in the activation of the chamber wall, composed by the ferritic steel HT-9, is demonstrated by using an intermediate graphite reflector. A good activation performance in the chamber is demonstrated allowing SLB and different recycling options. Hypothetical environmental releases of 1 kg of target-emissions-facing SiC tubes and HT-9 materials have also been simulated indicating optimum performances. (orig.)

  10. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    International Nuclear Information System (INIS)

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given

  11. Student internship program using the experimental fast reactor Joyo and related facilities

    International Nuclear Information System (INIS)

    The student training courses using the experimental fast reactor Joyo of the Japan Atomic Energy Agency (JAEA) and related facilities have been initiated based on the JAEA's mission to contribute to the human resources development program of the Japanese Ministry of Education, Culture, Sports, Science and Technology (MEXT) and Ministry of Economy, Trade and Industry (METI). The development of the student training courses was also strongly supported by the faculty of nuclear engineering of domestic universities in two reasons: one is that the nuclear related curriculum has recently been reduced due to the trend of decreasing interest by the younger generation in nuclear research and industry, and the other reason is that the aging research reactors and nuclear facilities owned by the universities are very difficult to keep operating. Considering this situation, JAEA decided to cooperate with the universities in developing the student training course. The experimental fast reactor Joyo of JAEA is a sodium cooled fast reactor with plutonium- uranium mixed oxide (MOX) fuel, which has two primary sodium loops, two secondary loops, and an auxiliary system. An intermediate heat exchanger (IHX) separates radioactive sodium in the primary system from non-radioactive sodium in the secondary system. The secondary sodium loop transports the reactor heat from the IHX to the air-cooled dump heat exchanger (DHX). Joyo has a full-scope type core and plant simulator, which duplicated all the main control panels located in the Joyo central control room. The simulator enables to offer a real time simulation of the plant behaviors under normal and abnormal conditions by applying the plant dynamic analysis code Minir-N2 and the same interlock system as the Joyo reactor system. The sodium analysis facility is located apart from Joyo complex to primarily conduct impurity measurement of Joyo cooling system. These data were measured by chemical analysis, gas chromatography, beta

  12. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 1014 cm-3.s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  13. Study on external dose around the Reactor TRIGA PUSPATI (RTP) Facility: A proposal

    International Nuclear Information System (INIS)

    In order to meet the requirement of the recent regulation (AELB-BSRP 2010), it is absolutely necessary to re-execute the in-situ and accumulated external dose assessment at the surrounding area of the Reactor TRIGA PUSPATI (RTP) facility. A number of strategic locations will be identified for the points of the dose mapping. Selection of these measurement points will be base on certain factor such as physical shielding thickness, occupancy of the workers, and others. Then, several survey meters and dosimeter will be chosen base on measuring method, reactor radiation spectra energy, type of radiation and etc. The result obtained will be compared with action values or limits given by AELB- BSRP 2010 and also can be used as baseline data or report for future reference. (author)

  14. Calculational framework for safety analyses of non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

  15. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10B (n, α) 7Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  16. Corrosion of aluminium-clad spent fuel in LVR-15 Research Reactor storage facilities

    International Nuclear Information System (INIS)

    This report documents the work performed under the IAEA Coordinated Research Project (CRP) on corrosion of the research reactor aluminium-clad spent fuel in water in the Nuclear Research Institute Rez. The aim of the project was to evaluate the corrosion of coupons of aluminium alloys used as cladding material of research reactor fuel elements, upon exposure to the water on the spent fuel storage basins. The corrosion of coupons exposed to two storage facilities at our Institute was investigated. Test racks were delivered by the IAEA and these contained coupons of two aluminium alloys, AA 6061 and SZAV-1. The racks also contained bimetallic couples consisting of aluminium alloy and stainless steel coupons. Rolled and extruded AA 6061 coupons were also tested. The single coupons, bimetallic couples and coupons with crevice couples were immersed in the at-reactor basin (ARB) and in the high-level waste pool (HLW). The chemical parameters of water in the two storage facilities were monitored and the extent of sedimentation of solids was measured. The ionic impurities were mainly Cl- and SO42- and their contents were 2 -15 μg/l in the HLW pool and about 20-250 μg/l in ARB. The iron content was below 2 μg/l in both facilities. After two years of exposure, pitting of the coupons was evaluated. Pits were observed mainly on the surfaces of single coupons and on the outer and inner surfaces of bimetallic and crevices coupons. No correlation was found between pitting and the type of aluminium alloy or between rolled and extruded materials. In the bimetallic couples, contact with stainless steel coupons did not have any affect on localized corrosion of the Al coupons. The pit depths were less than 50 μm on most of the coupon surfaces. Data obtained at this Institute should be compared with the results of other participants of this CRP. (author)

  17. Gamma exposure rate estimation in irradiation facilities of nuclear research reactors

    International Nuclear Information System (INIS)

    There are experimental situations in the nuclear field, in which dose estimations due to energy-dependent radiation fields are required. Nuclear research reactors provide such fields under normal operation or due to radioactive disintegration of fission products and structural materials activation. In such situations, it is necessary to know the exposure rate of gamma radiation the different materials under experimentation are subject to. Detectors of delayed reading are usually used for this purpose. Direct evaluation methods using portable monitors are not always possible, because in some facilities the entrance with such devices is often impracticable and also unsafe. Besides, these devices only provide information of the place where the measurement was performed, but not of temporal and spatial fluctuations the radiation fields could have. In this work a direct evaluation method was developed for the 'in-situ' gamma exposure rate for the irradiation facilities of the RA-1 reactor. This method is also applicable in any similar installation, and may be complemented by delayed evaluations without problem. On the other hand, it is well known that the residual effect of radiation modifies some properties of the organic materials used in reactors, such as density, colour, viscosity, oxidation level, among others. In such cases, a correct dosimetric evaluation enables in service estimation of material duration with preserved properties. This evaluation is for instance useful when applied to lubricating oils for the primary circuit pumps in nuclear power plants, thus minimizing waste generation. In this work the necessary elements required to estimate in-situ time and space integrated dose are also established for a gamma irradiated sample in an irradiation channel of a nuclear facility with zero neutron flux. (author)

  18. The 30 kW research reactor facility in Ghana: Past, present and future programmes

    International Nuclear Information System (INIS)

    The Ghana Research Reactor-1 (GHARR-1) is a small, simple, reliable and safe reactor design and constructed by China Institute of Atomic Energy (CIAE). GHARR-1 adopts the pool-tank structure and employs highly enriched uranium as fuel, light water as moderator and coolant, metal beryllium as reflectors. The reactor is cooled by natural convention. The rated maximum thermal power of GHARR-1 is 30 kW; the corresponding neutron flux is 1.0x1012 cm-2s-1. The refueling mode of the reactor is to totally change the old core with a new one, the lifetime being more than ten years. Since the commencement of operation of the low-flux miniature neutron source reactor (MNSR) in 1995, a significant number of research and development in the field of neutron activation analysis have taken place. During its 12 years of operation, after the first criticality, the reactor has been used as a neutron source for research, teaching and training to support several graduate and post graduate careers for students from universities in Ghana and the West African sub-region. Owing to the stable flux of the reactor and rapid proliferation in utilization, several analytical techniques have been developed. As a national neutron source reactor facility, Ghana's MNSR also known as GHARR-1 is now successfully utilized in various areas of neutron activation analysis (NAA), teaching, research and training. The GHARR-1 application in neutron activation analysis included: (i) Food analysis; (ii) Heavy metals determination in environmental samples; (iii) Determination of major, minor and trace elements in geological samples; (iv) And mineral prospecting among others. The educational programmes in place at the center are teaching and learning in nuclear engineering, nuclear physics, nuclear and radiochemistry and other related fields. The paper will focus on the past and current status of GHARR-1 with respect to utilization and management and future programmes to enhance its uses in the fields of teaching

  19. Consideration on rationalization of reactor safety systems. Evaluation of impact to plant safety and operations against facility sharing

    International Nuclear Information System (INIS)

    In the Feasibility Study of JNC, sodium cooled large-scale fast reactor and medium-scale modular reactor are studied as two of promising concepts. Sodium cooled medium-scale modular reactor has disadvantage compared to large-scale reactor that has large-scale effect. To compensate the disadvantage, modular effects are facilitated to improve facility sharing among modules, learning effect and so on. But facility sharing makes the plant system complex and impacts to plant safety and operations are worried. We evaluated the impacts to plant safety and operations by facility sharing in sodium cooled medium-scale modular reactor. We arranged ideas of facility sharing and extracted four ideas which are considered to have big impacts to safety or operations. 1) Sharing of the central control room, 2) Reduction of emergency generators, 3) Adoption of large capacity turbines, 4) Sharing of auxiliary cooling water system. In this study, we evaluated impacts to safety, operability, annual inspection and availability under consideration of the risk for unscheduled shutdown against above four facility sharing plans. Consistency of the facility sharing was confirmed. (author)

  20. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  1. Biological dosimetry studies for boron neutron capture therapy at the RA-1 research reactor facility

    International Nuclear Information System (INIS)

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminescent dosimeters to characterize the BNCT facility developed at the RA-1 research reactor operated by the National Atomic Energy Commission in Buenos Aires. Biological dosimetry was performed employing the hamster cheek pouch oral cancer model previously validated for BNCT studies by our group. Results indicate that the RA-1 neutron source produces useful dose rates for BNCT studies but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications. (author)

  2. Irradiation behaviour of TRIGA-LEU fuel in the TRIGA 14 MW reactor facility

    International Nuclear Information System (INIS)

    In order to convert TRIGA reactors to low enriched uranium fuel cycle, General Atomic (USA) produces a new type of fuel elements (U 235 enrichment -19.7%). A part of the TRIGA 14 MW (th) core in Pitesti consists of this type of fuel, and six elements have been examined in our post-irradiation facility. These were the first measurements using non-destructive control on this type of fuel. In this paper, some interesting information about the fuel behaviour during irradiation, is presented. (Author)

  3. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford Site

    International Nuclear Information System (INIS)

    This paper presents the methods used during the walkdowns to compile as-built structural information and analysis strategy to seismically qualify or verify the seismic adequacy of the safety class 1 components in the Plutonium Finishing Plant complex to current seismic criteria. Safety class 1 components and systems are those that perform a function required for nuclear criticality safety or whose failure might result in a significant release of radioactive, hazardous, or toxic materials as defined by DOE Order 6430.1A(1). The Plutonium Finishing Plant is a non-reactor nuclear facility designed and built during the 1950s to the Uniform Building Code, approximately 0.1g seismic requirement. The facility is located at the U.S. Department of Energy Hanford Site near Richland, Washington

  4. Utilization and facility of neutron activation analysis in HANARO research reactor

    International Nuclear Information System (INIS)

    The facilities of neutron activation analysis within a multi-purpose research reactor (HANARO) are described and the main applications of Neutron activation analysis (NAA) in Korea are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system, are installed at three irradiation holes. One irradiation hole is lined with a cadmium tube for epithermal-nal NAA. The performance of the NAA facility was examined to identify the characteristics of tube transfer system, irradiation sites and polyethylene irradiation capsule. The available thermal neutron flux with each irradiation site are in the range of 3.9x1013-1.6x1014 n/cm2·s and cadmium ratios are 15-250. Neutron activation analysis has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials and various polymers for research and development. Analytical services and the latest analytical results are summarized. (author)

  5. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    International Nuclear Information System (INIS)

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory's (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency's Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  6. Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non-time critical removal

  7. Life Management Programme for Long Term Operation of Reactors and Nuclear Facilities. Ageing Management of Research Reactors in Argentina

    International Nuclear Information System (INIS)

    The reactor RA-0 is a critical facility for the performance of exercises for research, education and training. It is located in the city of Cordoba, in the building of the Faculty of Physical Sciences. Its rated power is 1 W, which minimizes the shielding requirements for civil work and dispenses with a core cooling system. The core consists of two concentric and removable tanks of anodized aluminium with an active volume of 70 l. Moderator is demineralized light water, which enters through the bottom of the external tank with an overflow at the top of the inner tank that is removable. The fuel elements of 20.00% enriched UO2 are housed vertically in a grid with 232 holes of diameter greater than the fuel cladding. Other holes of smaller diameter allow circulation of the moderator, and four holes accommodate detector tubes or tubes for mounting experience. The control system consists of four control rods built with a cadmium sheet wrapped in a stainless steel cladding. These are inserted vertically and tangentially to the outer tank. Each bar is magnetically coupled to an electromagnet secured to the rise and fall mechanism, which also allows them to function as control rods. In case of emergency the electromagnet stops, and the bar falls under the influence of gravity. Obsolescence was detected in some elements of instrumentation, notably in former relays, starting systems and scram functions, but checks for updates in the data acquisition system have also been undertaken proactively to keep updated to the reactor. No ageing in mechanical systems has been verified

  8. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1988

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1988. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1988 annual reports submitted by about 429 licensees indicated that approximately 220,048 individuals were monitored, 113,00 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.20 rem (cSv) and an average measurable dose of 0.41 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,072 individuals completed their employment with one or more of the 429 covered licensees during 1988. Some 80,211 of these individuals terminated from power reactor facilities, and about 8,760 of them were considered to be transient workers who received an average dose of 0.27 rem (cSv). 17 refs., 11 figs., 29 tabs

  9. VEERA facility for studies of nuclear safety in VVER type reactors

    International Nuclear Information System (INIS)

    The VEERA facility was built in 1987 for experiments that simulate soluble neutron poison (boric acid) behaviour in a pressurized water reactor (PWR) during the long-term cooling period of loss-of-coolant accidents (LOCAs). The experiments provided insight especially into the processes of concentration, mixing and possible crystallization of boric acid in the core region of a PWR. In 1993 the facility was modified in order to use it for studies of the reflooding phenomenon. The results of the reflood experiments will be used as a data base for testing the capability of the reflood models of different computer codes. The VEERA facility in its original and modified forms is described in this report. Details of the geometry and dimensions of the components are given. This data is needed as a geometrical boundary condition in input deck preparation for thermal hydraulic analysis. The instrumentation and the data acquisition system are described so that the applicability of the facility and the accuracy of the measurements for different types of experiments can be evaluated. Initial and boundary conditions of the experiments and the principal test procedures are also summarized. (orig.) (24 figs., 6 tabs.)

  10. Waste inventory report for reactor and fuel-fabrication facility wastes

    International Nuclear Information System (INIS)

    The physical and chemical characteristics of wastes (other than spent fuel) that are generated at light-water-cooled nuclear reactor (LWR) power plants and nuclear fuel-fabrication facilities are described. The information used to define these characteristics is based on a survey of 30 nuclear power plants and 7 facilities involved in LWR fuel-fabrication. Information on volumes and activity levels of five major categories of LWR and fuel fabrication facility waste was collected. The major categories of waste are spent resin, concentrated liquids, precoat filter sludge (including ground ion-exchange resin), cartridge filters, and compactible and non-compactible trash. The volumes and activity levels of these wastes were used to determine the average waste-generation rates for LWRs and fuel-fabrication facilities, and the typical waste-generation rates for LWRs. Average gross radioactive concentrations were calculated, including concentrations of transuranics, to determine annual activity-generation rates and the total radioactivity of the waste shipped. These generation rates were used in conjunction with projected increases in the gross electrical generating capacity of nuclear power plants to determine annual waste volumes through 2000. A determination of the effect of a broad application of current volume-reduction systems on the waste volumes is also included. The overall waste burial capacity of currently licensed burial site areas was examined and fuel-cycle wastes and nonfuel-cycle wastes were considered

  11. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1989

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC 1 licensees during the years 1969 through 1989. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC 1 licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1989 annual reports submitted by about 448 licensees indicated that approximately 216,294 individuals were monitored 111,000 of whom were monitored by nuclear power facilities. They incurred an average individual does of 0.18 rem (cSv) and an average measurable dose of 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,535 individuals completed their employment with one or more of the 448 covered licensees during 1989. Some 76,561 of these individuals terminated from power reactor facilities, and about 10, 344 of them were considered to be transient workers who received an average dose of 0.64 rem (cSv)

  12. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1991

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1991. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1991 annual reports submitted by about 436 licensees indicated that approximately 206,732 individuals were monitored, 182,334 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.15 rem (cSv) and an average measurable dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 96,231 individuals completed their employment with one or more of the 436 covered licensees during 1991. Some 68,115 of these individuals terminated from power reactor facilities, and about 7,763 of them were considered to be transient workers who received an average dose of 0.52 rem (cSv)

  13. Long-term storage facility for reactor compartments in Sayda Bay - German support for utilization of nuclear submarines in Russia

    International Nuclear Information System (INIS)

    The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted waste monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on-shore concrete

  14. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  15. Power burst reactor facility as an epithermal neutron source for brain cancer therapy

    International Nuclear Information System (INIS)

    The Power Burst Facility (PBF) reactor is considered for modification to provide an intense, clean source of intermediate-energy (epithermal) neutrons desirable for clinical studies of neutron capture therapy (NCT) for malignant tumors. The modifications include partial replacement of the reflector, installation of a neutron-moderating, shifting region, addition shielding, and penetration of the present concrete shield with a collimating and (optionally) filtering region. The studies have indicated that the reactor, after these modifications, will be safely operable at full power (28 MW) within the acceptable limits of the plant protection systems. The neutron beam existing from the collimator port is predicted to be of sufficient intensity (∼ 1010) neutrons/cm2-s) to provide therapeutic doses in very short irradiation times. The beam would be relatively free of undesirable fast neutrons, thermal neutrons and gamma rays. The calculated neutron energy spectrum and associated gamma rays in the beam were provided as input in simulation studies that used a computer model of a patient with a brain tumor to determine predicted dose rates to the tumor and healthy tissue. The results of this conceptual study indicate an intense, clean beam of epithermal neutrons for NCT clinical trials is attainable in the PBF facility with properly engineered design modifications. 9 references, 11 figures, 3 tables

  16. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    International Nuclear Information System (INIS)

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term

  17. Conceptual design analysis of 4 K irradiation facility in Korean HANARO research reactor

    International Nuclear Information System (INIS)

    A conceptual design of a 4 K irradiation test facility has been conducted in support of the International Thermonuclear Experimental Reactor (ITER) magnet development program. A new research reactor designated as HANARO at the Korea Atomic Energy Research Institute has a Cold Neutron Source (CNS) port that is identified to be suitable for the fast neutron irradiation of metals and insulation materials for superconducting magnets at 4 K. A 40 hours of irradiation at full power will produce 2.5 x 1017 n/cm2 of the ITER magnet design neutron fluence with energy above 0.1 MeV. A material testing laboratory of Irradiated Materials Evaluation Facility (IMEF) that is located next to HANARO has been equipped with 77 K test machines and fracture analysis microscopes for radioactive specimens which can be upgraded for 4 K test without any intermediate warming. CNS radiation spectrum determined by Monte Carlo method is found to be more favorable for metal irradiation than for insulation materials with absorbed gamma dose that is 7-10 times the fast neutron dose. A lead-shielded irradiation capsule design with a 1 cm diameter specimen in 3 cm cold-bore diameter and 18 cm height will require about 120 watt cooling capacity at 4.6 K

  18. The DRAGON aerosol research facility to study aerosol behaviour for reactor safety applications

    International Nuclear Information System (INIS)

    During a severe accident in a nuclear power plant fission products are expected to be released in form of aerosol particles and droplets. To study the behaviour of safety relevant reactor components under aerosol loads and prototypical severe accident conditions the multi-purpose aerosol generation facility DRAGON is used since 1994 for several projects. DRAGON can generate aerosol particles by the evaporation-condensation technique using a plasma torch system, fluidized bed and atomization of particles suspended in a liquid. Soluble, hygroscopic aerosol (i.e. CsOH) and insoluble aerosol particles (i.e. SnO2, TiO2) or mixtures of them can be used. DRAGON uses state-of-the-art thermal-hydraulic, data acquisition and aerosol measurement techniques and is mainly composed of a mixing chamber, the plasma torch system, a steam generator, nitrogen gas and compressed air delivery systems, several aerosol delivery piping, gas heaters and several auxiliary systems to provide vacuum, coolant and off-gas treatment. The facility can be operated at system pressure of 5 bars, temperatures of 300 deg. C, flow rates of non-condensable gas of 900 kg/h and steam of 270 kg/h, respectively. A test section under investigation is attached to DRAGON. The paper summarizes and demonstrates with the help of two project examples the capabilities of DRAGON for reactor safety studies. (authors)

  19. Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP

    International Nuclear Information System (INIS)

    The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. - Highlights: • An epithermal neutron irradiation facility modelled using MCNPX. • Foils and TLDs used to measure dose in chamber and compared to simulations. • Proposed modifications to the irradiation chamber outlined based upon results of simulations

  20. Utilization of Mtr analysis of neutronic parameters for the neutron transmutation doping facility in the egyptian second research reactor

    International Nuclear Information System (INIS)

    since the early 1970's, semiconductor manufacturers considered using research reactor thermal neutrons to dope silicon with phosphorus. the method can result in a product with an inhomogeneity of 5% or less compared with conventional methods where the homogeneity fluctuated by a factor of about tow across the crystal. the greatly improved method of doping has resulted in a growing number of research reactor facilities engaged in commercial production of doped silicon. one of the very promising facilities of the ETRR-2 is the neutron transmutation - doping (NTD) unit, which is currently under commissioning. this work is done to optimize the performance of the ETRR-2 NTD facility based on the standard sand to develop a calculation line to be followed during its commercial operating phase. first, a review of the neutron transmutation doping techniques and description of the facilities around the world presented together with a description of the ETRR-2 facility

  1. Status of spent fuel in the 3MW BAEC MK-II research reactor facility of Bangladesh

    International Nuclear Information System (INIS)

    Bangladesh has been operating a 3 MW TRIGA MARK II research reactor since 1986. The reactor is installed in the campus of the Atomic Energy Research Establishment (AERE) at Savar, which is located about 40 km northwest of Dhaka. It is one of the main nuclear research facilities in the country. The reactor uses TRIGA LEU fuel with uranium content of 20% by weight. The enrichment level of the fuel is 19.7%. So far the reactor has been operated for 5624 hours with a total cumulative burnup (BU) of 10 690 MWh (445 MWd). The main areas of use are: training of man-power for research reactor operation and applications, radioisotope (RI) production, neutron activation analysis (NAA), neutron radiography (NR) and neutron scattering. Radioisotopes produced to date are: I-131, Sc-46 and Tc-99m. Bangladesh is a peace loving country with a strong commitment towards nuclear nonproliferation. Accordingly, it has signed several multilateral and bilateral agreements, protocols, treaties, etc. prevailing in the International Nuclear Non-proliferation regime. Bangladesh has also signed a Nuclear Cooperation Agreement with the USA on 17 September 1981, which facilitated export of nuclear technology from the USA to Bangladesh. The research reactor was procured under the provisions of this agreement. In 2003, the tenure of the Agreement was extended up to 2012. At present, there does not exist any spent fuel element in the reactor facility. However, with the recently undertaken RI production enhancement program, it is expected that the reactor will start generating spent fuels from the year 2012. It is to be mentioned that Bangladesh is aware of the US DOE's 'Take Back Program' in connection with the research reactor spent fuel of US origin, and is very much interested to take part in this program. The paper presents the current status of handling and storage facilities available for spent fuel and strategy for the safe management spent fuel to be generated from the research reactor in

  2. Loop facility for investigating of BRIG-300 reactor fuel elements with dissociating N2O4 coolant on ''Mariya'' reactor (PPR)

    International Nuclear Information System (INIS)

    The construction of a gas-cooled loop facility with a dissociating coolant N2O4 is designed that allows the fuel element to be tested under the simultaneous effect of the reactor irradiation and chemically active N2O4 with fast neutron density of (4-6)x1014 neutr./(cm2xs) and pressure of 16.0 MPa. The schematic drawing of the facility and loop channel construction are presented. The principle of the facility operation is described

  3. State of exposure control for workers engaging in radiation works and state of radioactive waste management in nuclear reactor facilities for test and research and nuclear reactor facilities at research and development stage, fiscal year 1995

    International Nuclear Information System (INIS)

    This is the summary of the reports submitted in fiscal year 1995 by the installers of the nuclear reactor facilities for test and research or at research and development stage, conforming to the related law. The individual dose equivalent of the workers engaging in radiation works in fiscal year 1995 was sufficiently lower than the prescribed limit in all reactor facilities. As for the released quantities of gaseous and liquid wastes, the radioactive substances in the air and water outside the monitor zones never exceeded the prescribed concentration limit in all reactor facilities. In the reactor facilities, for which the target values of release control have been determined, the values were less than the targets in all cases. The increase of stored radioactive solid waste decreased as the dismantling works of the reactor auxiliary system of the nuclear powered ship 'Mutsu' were finished in fiscal year 1994. As the amount of stored radioactive solid waste approaches the installed capacity, the preservation capacity of the existing waste preservation building was increased. (K.I.)

  4. Underwater plasma arc cutting of in-reactor tube of In-Pile Creep Test Facility

    International Nuclear Information System (INIS)

    The in-reactor tube of the In-Pile Creep Facility had been irradiated periodically for over 6 years in the Japan Materials Testing Reactor (JMTR) up to the end of 1978 under an operating condition of high temperature and high pressure identical to that of the Prototype Advanced Thermal Reactor, FUGEN, to gain the basic data for estimating the amount of creep which would occur in the pressure tubes of FUGEN. Following the removal of the in-reactor tube out of the JMTR, the test sections in the tube which were to be subjected to post irradiation examination were cut out. Underwater plasma arc cutting was employed to prevent the spread of contamination to the work area, to confine the heat affected zone in the test pieces to a minimum and to simplify disposal of the unneeded portions of the pressure tube. Setup of the cutting machine, cutting operations, radiological conditions during cutting of the highly radioactive portion of the tube and disassembly of the cutting equipment are described. In addition a brief description of the underwater plasma arc cutting machine is presented. The hot-cutting operations were done remotely to control personal exposure. The containment envelope prevented the spread of contamination to the environment and radioactive particles deposited on the cutting machine were removed without any difficulties. External exposure received by cutting personnel was small. Internal radionuclide deposit examinations were conducted, determining no crew member inhaled radioactive substances. Contamination spreads to the work area were minimal and release of radionuclide was well controlled. (author)

  5. Test Results of Reactor Coolant System Natural Circulation using the SMART-ITL Facility

    International Nuclear Information System (INIS)

    In this paper, the Sequence Of Event (SOE) and test conditions of RCS natural circulation test using SMART-ITL are presented, and the major measuring parameters and the test results will be introduced. In this test, the steady state operation satisfied the initial condition of the prescribed test procedure and the boundary conditions were properly simulated. After the RCPs stop, the RCS natural circulation flow was generated by heating in the core region and cooling in the SG heat exchanger region, and the major thermalhydraulic parameters reached at a stable condition. Through this experiment, it has been validated that the SMART-ITL facility can adequately simulate the RCS natural circulation behavior. In addition, it is expected that the experimental data can be used for the code assessment of the TASS/SMR-S code and experiences from this test can be utilized to the subsequent SBLOCA simulation test. SMART (System-integrated Modular Advanced Reactor) is an integral type reactor which major primary components such as the steam generator, the pressurizer, and the reactor coolant pump are installed inside one single reactor vessel and connecting primary pipes are removed. The TASS/SMR-S code is used to perform the performance and safety analysis of the SMART. To evaluate the capability of TASS/SMR-S code on the natural circulation and accident scenarios such as Small-Break Loss of Coolant Accident (SBLOCA) for predicting the thermal-hydraulic phenomena in steady state and transient operation, it is essential to perform a series of validation tests

  6. Training and accreditation activities at the Department of Energy Category A Reactors and Nuclear Facilities

    International Nuclear Information System (INIS)

    A new era dawned within the Department of Energy (DOE) in 1989 when DOE Order 5480.18A 'Accreditation of Performance-Based Training for Category A Reactors and Nuclear Facilities' was issued. This new era emphasized the importance of personnel training and qualification in maintaining the continued safe and efficient operation of the diverse nuclear facilities within the DOE complex. This approach to the design, development, and implementation of training is very similar to the approach that has proven to be very successful within the commercial nuclear industry. During the 1980s in the aftermath of Three Mile Island (TMI), DOE made a significant effort to conduct its mission in an environmentally safe manner and to increase the existing level of protection of the workers and the public. The DOE, like the commercial nuclear industry, realized that a nuclear accident anywhere in the U.S.A. would negatively impact the public confidence in the entire industry. This effort has not been easy because of changes within DOE and changes in regulatory requirements. Difficulties include aging facilities, outdated equipment, ingrained operating habits, inflexible culture, outdated or nonexistent procedures, stagnated management structure, lack of technical specifications and safety analysis reports, no configuration control, informal and undocumented training and qualification processes, and a wide diversity of operating facilities. The commercial industry had problems adjusting to the new regulations after TMI, but none as challenging as those facing DOE. This paper centers on the importance and status of accreditation within the DOE community and the efforts to develop and implement a performance-based approach to the training of the personnel at these facilities

  7. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    International Nuclear Information System (INIS)

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs

  8. Estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities: a summary

    International Nuclear Information System (INIS)

    A project has been carried out at Oak Ridge National Laboratory to assess the human health and environmental effects associated with the operation of a liquid-metal fast breeder reactor fuel cycle. In this first phase of the work, emphasis was on routine radionuclide releases from reactor and reprocessing facilities. Sites for 51 1-GW(e) capacity reactors and 3 reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and the reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also the results of these analyses are being used in the identification of research priorities

  9. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  10. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G.; Estrade, J. [French Atomic Energy Commission, Paris (France)

    2013-07-01

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  11. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    International Nuclear Information System (INIS)

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described

  12. Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

  13. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  14. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  15. Performance testing of irradiation facility rabbit system pneumatic reactor RSG-GAS using standard reference material

    International Nuclear Information System (INIS)

    The irradiation facility function test of rabbit system pneumatic (RS-5) has been done using standard reference material SRM 1633 Coal Fly ash through the sending station. Long irradiation of about 4-5 seconds. The results of qualitative analysis showed that the dominant elements listed in the certificate can be detected are Al, Ca, Mg, Si, Na, Ti, V, Mn. But only an element of Mn and Na which has a relative refractive values below 10%. And the other elements have a value relative refractive index 25% - 60%. The significant difference of value was not influenced by the position of irradiation in the reactor facility but due to the influence of the time difference between the sample and the standard count, and the half-life nuclide itself. Overall it can be said that the performance of the irradiation facility pneumatic rabbit system is good, but needs to be tested again by using different standard reference materials, in order to obtain the test results of analysis that can be trusted. (author)

  16. Ventilation and air cleaning plant experience in fast reactor fuel cycle facilities at Dounreay

    International Nuclear Information System (INIS)

    This paper is about work to measure the quantity and quality of aerosols created in Fast Reactor Fuel Cycle Facilities. Such facilities typically contain large amounts of Pu, Actinides, Fission Products often in a dispersable form, which pose varied challenges to ventilation systems in general and clean up devices in particular. Fluidic (no moving part) devices have been used intensively for the past 5 years at Dounreay and their performance will be discussed. Of particular importance has been their use on small enclosures such as gloveboxes, where they have made an important contribution in minimising activity escape. A large amount of data on HEPA filter performance in various parts of the facility has now been accumulated. Testing experience, particularly in situ of HEPAs will be presented and the problems encountered discussed. Factors affecting HEPA life will be discussed. In the past the development of new filter housing and changing systems have been presented to the US ERDA Air Cleaning Conference. The performance of these devices under active conditions will be reported. Considerable work has been done on the estimation of residual alpha activity on used HEPAs by neutron counting and equipment developed to allow alpha activity to be detected down to very low levels. This can also be done for high gamma active filters inside lead shielding. System failures are discussed together with performance of safety equipment. (author)

  17. Characterization and adjustment of the neutron radiography facility of the RP-10 nuclear reactor

    CERN Document Server

    Ravello-R, Y R

    2001-01-01

    The main aim of this work was to characterize and adjust the neutron radiography facility of the RP-10 nuclear reactor, and therefore be able to offer with this technique services to the industry and research centers in general. This technique will be complemented with others such as x-rays and gamma radiography. First, the shielding capacity of the facility was analyzed, proving that it complies with the radiological safety requirements established by the radiological safety code. Then gamma filtration tests were conducted in order to implement the direct method for image formation, optical density curves were built according to the thickness of the gamma filter, the type of film and the type of irradiation. Also, the indirect method for image formation was implemented for two types of converters: indium and dysprosium. Growth curves for optical density were also made according to contact time between converter-film, for different types of films. The resolution of the facility was also analyzed using two met...

  18. Prompt gamma neutron activation analysis facility at the RA-6 research reactor

    International Nuclear Information System (INIS)

    A prompt gamma neutron activation activation analysis facility was developed at the 500 kw thermal power RA-6 research reactor of the Bariloche Atomic Center, Argentina.This facility consist of a radial beam port with external positioning of the sample.The gamma radiation is reduced by a bismuth filter placed inside the extraction tube and the beam diameter is limited by a set of two collimators up to 5 cm.The neutron flux at the sample position is 7 106 n/cm2s with a Cadmium ratio of 20/1.The gamma detector is a 50 % efficiency type p HPGe rounded by a NaI(Tl) for Compton suppressioning.The gamma spectra is measured through 0 to 8.5 MeV.The background have counting rate of 350 cps without sample. In this work is shown the efficiency curve, the calculed sensibilities and the lower detection limits for B, Cd, Sm, Gd, H, Cl, Hg, Eu, Ti, Ag, Au, Mo. The RA-6's PGNAA facility is fully working, although the analytic capacity is under improvement

  19. Selection of away-from-reactor facilities for spent fuel storage. A guidebook

    International Nuclear Information System (INIS)

    This publication aims to provide information on the approaches and criteria that would have to be considered for the selection of away-from-reactor (AFR) type spent fuel storage facilities, needs for which have been growing in an increasing number of Member States producing nuclear power. The AFR facilities can be defined as a storage system functionally independent of the reactor operation providing the role of storage until a further destination such as a disposal) becomes available. Initially developed to provide additional storage space for spent fuel, some AFR storage options are now providing additional spaces for extended storage of spent fuel with a prospect for long term storage, which is becoming a progressive reality in an increasing number of Member States due to the continuing debate on issues associated with the endpoints for spent fuel management and consequent delays in the implementation of final steps, such as disposal. The importance of AFR facilities for storage of spent fuel has been recognized for several decades and addressed in various IAEA publications in the area of spent fuel management. The Guidebook on Spent Fuel Storage (Technical Reports Series No. 240 published in 1984 and revised in 1991) discusses factors to be considered in the evaluation of spent fuel storage options. A technical committee meeting (TCM) on Selection of Dry Spent Fuel Storage Technologies held in Tokyo in 1995 also deliberated on this issue. However, there has not been any stand-alone publication focusing on the topic of selection of AFR storage facilities. The selection of AFR storage facilities is in fact a critical step for the successful implementation of spent fuel management programmes, due to the long operational periods required for storage and fuel handling involved with the additional implication of subsequent penalties in reversing decisions or changing the option mid-stream especially after the construction of the facility. In such a context, the long

  20. Concerning status of radioactive waste management and exposure dose to workers at operating nuclear reactor facilities in 1987

    International Nuclear Information System (INIS)

    The Guidelines on Target Environmental Dose around Light Water Reactor Facilities for Power Generation specifies target waste release management levels required to meet the target environmental doses. All operators of operating light water reactor facilities for power generation are required to meet the target waste release management levels. Radioactive solid wastes are required to be stored in sealed drums in appropriate solid waste warehouses. The present not briefly summarizes the data given in the 1987 Report on Radiation Management submitted in accordance with the Law Concerning Management of Nuclear Material, Nuclear Fuel and Nuclear Reactor by the operators of operating nuclear reactor facilities for power generation, and the Report on Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. These reports have shown that the amount of radioactive gas and liquid wastes released in 1987 was below the target level. Data on radioactive gas and liquid wastes released in power generation plants with operating nuclear reactor facilities in 1978 and the subsequent years are tabulated. Data on radioactive solid wastes are also shown. (N.K.)

  1. Results of clinical and hygienic observations of labour conditions and the state of health of personnel at research reactor facilities

    International Nuclear Information System (INIS)

    Complex studies of operating conditions and health status of the personnel employed in research reactors or in critical assembly have been undertaken in order to define more precisely the regulations, requirements and norms of radiation safety to be applied during the designing and operation of various facilities. The main potential sources of external radiation, dose rates for gamma radiation, neutron fluxes, and maximum annual doses received by personnel under normal reactor operation have been determined. Clinical physiological studies have shown that the health of the personnel employed in nuclear reactors is satisfactory

  2. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Alam, S. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Zaman, M.A. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Islam, S.M.A. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Ahsan, M.H. (Inst. of Nuclear Science and Technology (INST), AERE, Savar, Dhaka (Bangladesh))

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work. (orig.)

  3. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    International Nuclear Information System (INIS)

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work. (orig.)

  4. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Science.gov (United States)

    Alam, Sabina; Zaman, M. A.; Islam, S. M. A.; Ahsan, M. H.

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work.

  5. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  6. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    International Nuclear Information System (INIS)

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated

  7. Fuels and materials research under the high neutron fluence using a fast reactor Joyo and post-irradiation examination facilities

    International Nuclear Information System (INIS)

    The experimental fast reactor Joyo at Oarai Research and Development Center (ORDC) of Japan Atomic Energy Agency (JAEA) is Japan's sodium-cooled fast reactor (FR). In 2003, this reactor's upgrade to the 140MWt MK-III core was completed to increase the irradiation testing capability. The MK-III core provides the fast neutron flux of 4.0x1015n/cm2s as an irradiation test bed for improving the fuels and material of FR in Japan. Three post-irradiation examination (PIE) facilities named FMF, MMF and AGF related to Joyo are in ORDC. Irradiated subassemblies and core components are carried into the FMF (Fuel Monitoring Facility) and conducted nondestructive examinations. Each subassembly is disassembled to conduct some destructive examinations and to prepare the fuel and material samples for further detailed examinations. Fuel samples are sent to the AGF (Alpha-Gamma Facility), and material samples are sent to the MMF (Materials Monitoring Facility). These overall and elaborate data provided by PIE contribute to investigate the irradiation effect and behavior of fuels and materials. This facility complex is indispensable to promote the R and D of FR in Japan. And, the function and technology of irradiation test and PIE enable to contribute to the R and D of innovative fission or fusion reactor material which will be required to use under the high neutron exposure. (author)

  8. Measurements of reactor-relevant electromagnetic effects with the FELIX facility

    International Nuclear Information System (INIS)

    In predicting the electromagnetic consequences of a plasma disruption in a tokamak reactor design, a two-dimensional electromagnetic model of the first wall, blanket, and shield (FWBS) system is typically used. The response to a decaying plasma current is then found to be dominated by a single eddy-current mode, with a single L/R time. Recent experiments with the Fusion ELectromagnetic Induction eXperiment (FELIX) facility at Argonne National Laboratory suggest that such modeling can be used to design against electromagnetic forces and torques, but only if a range of values is used for both tau, the plasma decay time, and tau0, the L/R time of the FWBS system

  9. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    International Nuclear Information System (INIS)

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable

  10. Commercial Light Water Reactor Tritium Extraction Facility. Geotechnical Summary report (U)

    International Nuclear Information System (INIS)

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  11. A microprocessor based monitoring system for a small nuclear reactor facility

    International Nuclear Information System (INIS)

    An inexpensive microprocessor based system has been designed and constructed for our 250 kilowatt TRIGA reactor facility. The system, which is beginning operational testing, can monitor on a continuous basis the status of up to 54 devices and maintain a record of events. These devices include fixed radiation monitors, pool water level trips, security alarms and an access control unit. In the latter case, the unit permits selection of different levels of access permission based on the time of day. The system can alert security and other personnel in the event of abnormalities. Because of the inclusion of this in the security system, special reliability and failure mode operation. The unit must also be simple to install, program and operate. (author)

  12. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D.J.; Brehm, J.R.

    1994-01-01

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.

  13. Regularities of hydraulic irregularities formation at the reactor facility collector system outlet

    International Nuclear Information System (INIS)

    Theoretical and experimental studies of hydrodynamics of flow paths of axisymmetrical collector systems (CS) of cylindrical and plane forms with different variants of coolant input and output have been carried out. The regularities of hydraulic irregularities generation in the CS output of WWER type reactor facility have been ascertained. It is shown that position of the maximal coolant velocity in the outlet of the grid coincides with position of maximal velocity in the core of the jet falling on it, and the value of the maximal average coolant velocity in grid openings is proportional to the value of average coolant velocity in jet falling on it. It is pointed out that in CS flow paths coolant flow has complex hydrodynamic structure and is determined mainly by ratio of dimensions and CS design, coolant flow in its flow path and ratio of pressure drop in CS input and output. The data obtained make it possible to optimize hydrodynamics of considered CS without expensive experiments

  14. Dosimetric characteristics of the thermal neutron beam facility for neutron capture therapy at Hanaro reactor

    International Nuclear Information System (INIS)

    The thermal neutron beam facility utilizing a typical tangential beam port for Neutron Capture Therapy was installed at the Hanaro, 30 MW multi-purpose research reactor. In order to determine the different dose components in phantoms irradiated with a mixed thermal neutron beam and gamma-ray for clinical applications, various techniques were applied including the use of activation foils, TLDs and ionization chambers. The water phantom was utilized in the measurement. The results of the measurement were compared with MCNP4B calculations. The thermal neutron fluxes were 1.02E9 and 6.07E8/cm2·s at 10 and 20 mm depth in water, respectively. The gamma-ray dose rate was 5.10 Gy/hr at 20 mm depth in water. The result of this study can be used as basic data for subsequent BNCT clinical application. (author)

  15. The new cold neutron research facility at the Budapest Research Reactor

    International Nuclear Information System (INIS)

    The new cold neutron research facility is routinely operated at the Budapest Neutron Centre since February 2001. At the 10 MW research reactor a liquid hydrogen cold neutron source (CNS) has been installed. The commissioning of the CNS has been followed by the replacement of the old neutron guides by a new supermirror guide system both for the in-pile and out-of pile part. The ensemble of the CNS and new guides provides an intensity gain of the order of 30-60. The cold neutron channel has a take-off for three beams. The first guide serves for a triple axis spectrometer and a prompt gamma activation analysis station. A small angle scattering spectrometer is installed on the middle guide, and a reflectometer is operated on the third one. (author)

  16. An update on the LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada, has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL)-extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR-2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  17. Daily tritium intakes by people living near a heavy-water research reactor facility: dosimetric significance

    International Nuclear Information System (INIS)

    We have estimated the relative daily intakes of tritiated water (HTO) and organically bound tritium (OBT), and have measured HTO-in-urine, in an adult population residing in the town of Deep River, Ontario, near a heavy-water research reactor facility at Chalk River. The daily intake of elevated levels of atmospheric tritium has been estimated from its concentration in environmental and biological samples, and various food items from a local tritium-monitoring program. Where the available data were inadequate, we used estimates generated by an environmental tritium-transfer model. From these data and estimates, we calculated a total daily tritium intake of about 55 Bq. Of this amount, 2.5 Bq is obtained from OBT-in-diet. Inhalation of HTO-in-air (15 Bq d-1) and HTO-in-drinking water (15 Bq d-1) accounts for more than half of the HTO intake. Skin absorption of HTO from air and bathing or swimming (for 30 min d-1) accounts for another 9 Bq d-1 and 0.1 Bq d-1, respectively. The remaining intake of HTO is from food as tissue-free water tritium. The International Commission on Radiological Protection's recommended two-compartment metabolic model for tritium predicts an equilibrium body burden of about 900 Bq from HTO (818 Bq) and OBT (83 Bq) in the body, which corresponds to an annual tritium dose of 0.41 μSv. The model-predicted urinary excretion of HTO (∼18 Bq L-1) agrees well with measured HTO-in-urine (range, 10-32 Bq L-1). The OBT dose contribution to the total tritium dose is about 16%. We conclude that for the people living near the Chalk River research reactor facility, the bulk of the tritium dose is due to HTO intake. (author)

  18. Assessment of radioactive wastes from a DCLL fusion reactor: Disposal in El Cabril facility

    International Nuclear Information System (INIS)

    Highlights: •Radwastes from a DCLL reactor have been assessed following IAEA classification. •Disposal in El Cabril facility has been studied. •9% of the total volume considered can be disposed in El Cabril. •Concrete-made biological shield can be managed through clearance. •Comparison with French regulation shows similar conclusions. -- Abstract: Under the Spanish Breeding Blanket Technology Programme TECNOFUS a conceptual design of a DCLL (Dual-Coolant Lithium–Lead) blanket-based reactor is being revised. The dually cooled breeding zone is composed of He/LiPb and SiC as material of the liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer) for the blanket and austenitic steel (SS316LN) for the vacuum vessel (VV) and the cryostat. In this work, radioactive wastes are assessed in order to determine if they can be disposed as low and intermediate level radioactive waste (LILW) in the Spanish near surface disposal facility of El Cabril. Also, unconditional clearance and recycling waste management options are studied. The neutron transport calculations have been performed with MCNPX code, while the ACAB code is used for calculations of the inventory of activation products and for activation analysis, in terms of waste management ratings for the options considered. Results show that the total amount of the cryostat can be disposed in El Cabril joined to the outer layer of both VV and channel inserts, whereas only concrete-made biological shield can be managed through clearance and none of the steels can be recycled. Those results are compared with those corresponding to French regulation, showing similar conclusions

  19. Electromagnetic conductivity and magnetic surveys for characterization of an abandoned nuclear reactor facility

    International Nuclear Information System (INIS)

    Electromagnetic conductivity and total field magnetic surveys were conducted over a 14 acre portion of a site used during the Manhattan Project for nuclear research and plutonium production. While the facility was decommissioned and abandoned in 1953, there is presently some concern related to the potential for movement of radioactively contaminated ground water and sediments through high permeability underground features, such as tunnels or pipes. In addition, a number of underground storage tanks were left in place when the site was abandoned. The objectives of the surveys were to enhance the available data base related to the locations of underground features. The electromagnetic results clearly defined the presence of steam lines, water mains, electrical utility lines, building foundations, underground storage tanks, and a tunnel. The magnetic results were ineffective in detecting elongated subsurface features and provided relatively poor definition of building foundations. However, the magnetic results did detect disposal areas and reactor remains that were poorly defined by the electromagnetic data. The differences in the two data sets demonstrate the need for both electromagnetic and magnetic surveys in site characterization studies. Thirteen of the excavated anomalies were associated with buried utility features that could serve as high permeability conduits for ground water or sediment movement. The deepest of these was a 6 inch water main at a depth of 6 feet. The remainder of the utilities were at depths of 4 feet or less. Three of the anomalies resulted from underground storage tanks that ranged in size from 1000 gallons to 5000 gallons. Seven of the anomalies resulted from general waste disposal areas. The use of historical documentation combined with electromagnetic and magnetic surveys provided a cost effective method for baseline characterization of the complex underground features associated with a nuclear reactor facility

  20. Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    Energy Technology Data Exchange (ETDEWEB)

    Lisa Harvego; Brion Bennett

    2011-11-01

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  1. Characterization Of Normalization Factor In TRIGA 2000 Bandung Reactor Pneumatic Facility for Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Neutron activation analysis using synthetic multielement comparators is prevalent method for multielement analysis. This comparison method has several limitations such as preparation of synthetic standard is time consuming and needs high cost. In order to overcome such difficulties, the use of normalization factor of sample geometry and irradiation position as well need to be done. The normalization factor is used to overcome flux inhomogeneity, so that the used of standard reference material can be minimized. In this research, characterization of normalization factor in pneumatic facility of TRIGA 2000 Bandung reactor, have been done. The determination was done for two sample positions (bottom and top) using polyethylene container. The average normalization factor at 60,30 and 15 second irradiation at 1500 k Watt for Cu sample gave values of 1.2848, 1.2908 and 1.3348 respectively. The effect of power reactor fluctuation on normalization factor was also studied. Fluctuation of power reactor under 2 % for sample position top and bottom gave deviation values of 3.1699% and 1.6238% respectively. The determination of normalization factor for Ti, I, V and AI reference standards have also been done. Normalization factor at 60 second irradiation at 1500 k Watt for Ti, I, V and AI reference standards gave mean values of 1.2554, 1.2066, 1.3625 and 1.2475 respectively. Normalization factor obtained of this research have a narrow range (<6.2%). The results obtained can be use in developing the NAA method, to minimize the spent of time, energy and cost

  2. Temporal variation of the neutron flux in the carousel facility of a TRIGA reactor

    International Nuclear Information System (INIS)

    In this work we focused on identifying quantitatively the temporal (time-dependent) variation of neutron flux in the carousel facility (CF) of TRIGA reactor at the 'Jozef Stefan' Institute (IJS) for core No. 176, set up in April 2002. The measurements are based on neutron detectors (ionisation chambers), which surround the graphite reflector of the reactor core. In principle, the variations of the neutron flux produce a systematic error in the results obtained by absolute or 'quasi' absolute measuring techniques (such as neutron activation analysis (NAA) by the k0-standardization method), which assume constant conditions during irradiation. The results of our study show that for typical irradiation of 20 hours in channels of the CF aligned in the direction of the ionisation chamber (safety channel) the time-dependent variation of the neutron flux is about 6-8%. In the k0 method, which we are using for routine work at the IJS, this variation introduced a systematic error in the results up to 4.6%, depending on the half-life of investigated radionuclide. (author)

  3. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    International Nuclear Information System (INIS)

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 1014 nominally at the nose in the D2O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  4. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  5. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    International Nuclear Information System (INIS)

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  6. Development and Operation of Experiment Course using Research Reactor and Associated Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Shin, B. C.; Hwang, I. A.; Won, J. Y.; Ju, Y. C.; Nam, J. S.; Seo, K. W.; Kim, H. N.

    2013-05-15

    The purpose of present research is to offer a specialized educational opportunity by developing specific curriculum for potential users, mainly university students majoring in related with nuclear engineering and radiation field, on site at KAERI, exploiting the diverse offering of HANARO and ancillary facilities. The specific items of this research accomplished are: First, Development of various curricula for specific research using HANARO and continuous operation of the developed curricula to provided university students with opportunities to use HANARO. Second, Continuous operation of research reactor related experimental training programs for university students in nuclear field to make contribution to cultivating specialists. Third, through the site experimental training for new coming nuclear engineering students, support future potential users to the nuclear research fields, as well as enlarge or broaden the base. Finally, it is hoped that these experiments broadens public awareness and acceptance of the present and potential future contribution of the reactor technology, there by bring positive impacts to policy making. As a whole, 108 students offered and 88 students from 6 universities have completed the course of the programs developed by this project. Also, 1 textbook and 1 teaching aid, a questionnaire have been developed to support the program.

  7. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang-Hee; Kang, Young-Hwan; Kuk, Il-Hiun [Korea Atomic Energy Research Institute, Taejon (Korea)

    2001-03-01

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 10{sup 14} nominally at the nose in the D{sub 2}O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  8. Dosimetric Implications of Atmospheric Dispersal of Tritium Near a Heavy-water Research Reactor Facility

    International Nuclear Information System (INIS)

    An estimate of the tritium dose to the public in the vicinity of the heavy water research reactor facility at AECL-Chalk River Laboratories, Ontario, Canada, has largely been accomplished from analyses on regularly-collected samples of air, precipitation, drinking water and foodstuffs (pasture, fruit, vegetables and milk) and environmental dose models. To increase the confidence with which public doses are calculated, tritium doses were estimated directly from the ratio of tritiated species in urine samples from members of the general public. Single cumulative 24 h urine samples from a few adults living in the vicinity of the heavy-water research reactor facility at Chalk River Laboratories, Canada were collected and analysed for tritiated water and organically bound tritium. The participants were from Ottawa (200 km east), Deep River (10 km west) and Chalk River Laboratories. Tritiated water concentrations in urine ranged from 6.5 Bq.l-1 for the Ottawa resident to 15.9 Bq.l-1 for the Deep River resident, and were comparable to the ambient levels of tritium-in-precipitation at their locations. The ultra-low levels of organically bound tritium in urine from these same individuals were measured by 3He-ingrowth mass spectrometry and were 0.06 Bq.l-1 (Ottawa) and 0.29 Bq.l-1 (Deep River). For Chalk River Laboratories workers, tritiated water concentrations in urine ranged from 32 Bq.l-1 to 9.2x104 Bq.l-1, depending on the ambient levels of tritium in their workplace. The organically bound tritium concentrations in urine from the same workers were between 0.08 Bq.l-1 and 350 Bq.l-1. With a model based on the ratio of tritiated water to organically bound tritium in urine, the estimated dose arising from organically bound tritium in the body for the Ottawa and Deep River residents was about 26% and 50%, respectively, of the body water tritium dose. The workers in a reactor building at Chalk River Laboratories has less than 10% dose contribution from organically bound

  9. IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility

    International Nuclear Information System (INIS)

    1 - Description: The RRR-SEG-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactor RRG-SEG (at RC Rossendorf / Germany) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in Note Technique SPRC/LEPh/93-230 (SEG) The SEG experiments are considered 'clean' integral experiments, simple and clear in geometry and well suited for calculation. In all SEG configurations only a few materials were used, most of these were standards. Due to the designed adjoint function (energy-independent or monotonously rising), the capture or scattering effect was dominant, convenient to check separately capture or scattering data. At first, analyses of the experiments have been performed in Rossendorf. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and

  10. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    International Nuclear Information System (INIS)

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  11. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    Energy Technology Data Exchange (ETDEWEB)

    Dan Ogden

    2014-10-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

  12. Optimization of the irradiation beam in the BNCT research facility at IEA-R1 reactor

    International Nuclear Information System (INIS)

    Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatment of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discriminated at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to perform studies in the area and it has a great potential for improvement for tailoring the irradiation field. (author)

  13. Large-scale Samples Irradiation Facility at the IBR-2 Reactor in Dubna

    CERN Document Server

    Cheplakov, A P; Golubyh, S M; Kaskanov, G Ya; Kulagin, E N; Kukhtin, V V; Luschikov, V I; Shabalin, E P; León-Florián, E; Leroy, C

    1998-01-01

    The irradiation facility at the beam line no.3 of the IBR-2 reactor of the Frank Laboratory for Neutron Physics is described. The facility is aimed at irradiation studies of various objects with area up to 800 cm$^2$ both at cryogenic and ambient temperatures. The energy spectra of neutrons are reconstructed by the method of threshold detector activation. The neutron fluence and $\\gamma$ dose rates are measured by means of alanine and thermoluminescent dosimeters. The boron carbide and lead filters or $(n/\\gamma)$ converter provide beams of different ratio of doses induced by neutrons and photons. For the lead filter, the flux of fast neutrons with energy more than 0.1 MeV is $1.4 \\cdot 10^{10}$ \\fln and the neutron dose is about 96\\% of the total radiation dose. For the $(n/\\gamma)$ converter, the $\\gamma$ dose rate is $\\sim$500 Gy h$^{-1}$ which is about 85\\% of the total dose. The radiation hardness tests of GaAs electronics and materials for the ATLAS detector to be put into operation at the Large Hadron ...

  14. The pneumatic carrier facility in Dhruva reactor: commissioning, characterization and utilization

    International Nuclear Information System (INIS)

    The 100 MWt power Dhruva research reactor, BARC is provided with pneumatic carrier facility (PCF) to carry out R and D work using short-lived (seconds to minutes) radioisotopes in the fields like neutron activation analysis (NAA) and nuclear fission. The samples are kept inside a high density polypropylene capsule (rabbit), which is pneumatically sent to the irradiation position in the core and retrieved after a preset time of irradiation. After the irradiation, radioactivity assay is carried out using high resolution gamma ray spectrometry with HPGe detector coupled to PC based MCA. The availability of high neutron flux (∼ 5 x 10 13 cm-2s-1 at 50 MWt power) and shorter retrieval time (∼5 seconds) make it possible to measure short-lived isotopes with enhanced sensitivity. This report describes the salient features of this facility, characterization of the neutron spectrum at this irradiation position and its utilization. The PCF is being extensively utilized for analytical applications using NAA as well as nuclear fission studies. A brief description of analysis of some samples of geological, environmental and biological origin, nuclear materials as well as reference materials is included in this report. Protocol and check list for carrying out PCF irradiations and gamma spectrometric assay are also given at the end of the report. (author)

  15. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  16. Assessment of gold flux monitor at irradiation facilities of MINT TRIGA MK II reactor

    International Nuclear Information System (INIS)

    Neutron source of MINTs TRIGA MK II reactor has been used for activation analysis for many years and neutron flux plays important role in activation of samples at various positions. Currently, two irradiation facilities namely the pneumatic transfer system and rotary rack are available to cater for short and long lived irradiation. Neutron flux variation for both irradiation facilities have been determined using gold wire and gold solution as flux monitor. However, the use of gold wire as flux monitor is costlier if compared to gold solution. The results from analysis of certified reference materials showed that gold solution as flux monitors yield satisfactory results and proved to safe cost on the purchasing of gold wire. Further experiment on self-shielding effects of gold solution at various concentrations has been carried out. This study is crucial in providing vital information on the suitable concentration for gold solution as flux monitor. In the near future, gold solution flux monitor will be applied for routine analysis and hence to improve the capability of the laboratory on neutron activation analysis. (Author)

  17. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    2012-11-15

    ... NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to... filing requirements of the NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals of NRC... COMMISSION The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No....

  18. The Text of the Agreement for the Application of Agency Safeguards to Four United States Reactor Facilities

    International Nuclear Information System (INIS)

    The text of the Agreement between the Agency and the United States of America for the application of Agency safeguards to four United States reactor facilities, which was signed on 30 March 1962 and will enter into force on 1 June 1962, is reproduced in this document for the information of all Members of the Agency

  19. Establishment of the Neutron Beam Research Facility at the OPAL Reactor

    International Nuclear Information System (INIS)

    Full text: Australia's first research reactor, HIFAR, reached criticality in January 1958. At that time Australia's main agenda was establishment of a nuclear power program. HIFAR operated for nearly 50 years, providing a firm foundation for the establishment of Australia's second generation research Reactor OPAL, which reached criticality in August 006. In HIFAR's early years a neutron beam facility was established for materials characterization, partly in aid of the nuclear energy agenda and partly in response to interest from Australia's scientific community. By the time Australia's nuclear energy program ceased (in the 1970s), radioisotope production and research had also been established at Lucas Heights. Also, by this time the neutron beam facility for scientific research had evolved into a major utilization programme, warranting establishment of an independent body to facilitate scientific access (the Australian Institute for Nuclear Science and Engineering). In HIFAR's lifetime, ANSTO established a radiopharmaceuticals service for the Australian medical community and NDT silicon production was also established and grew to maturity. So when time came to determine the strategy for nuclear research in Australia into the 21st century, it was clear that the replacement for HIFAR should be multipurpose, with major emphases on scientific applications of neutron beams and medical isotope production. With this strategy in mind, ANSTO set about to design and build OPAL with a world-class neutron beam facility, capable of supporting a large and diverse scientific research community. The establishment of the neutron beam facility became the mission of the Bragg Institute management team. This journey began in 1997 with establishment of a working budget, and reached its first major objective when OPAL reached 20 MW thermal power nearly one decade later (in 2006). The first neutron beam instruments began operation soon after (in 2007), and quickly proved themselves to be

  20. Utilisation and Upgrading of the Neutron Beam Lines Facilities at the SAFARI-1 Nuclear Research Reactor in South Africa

    International Nuclear Information System (INIS)

    The South African Nuclear Energy Corporation (Necsa) owns and operates the SAFARI-1 20 MW Research Reactor located near Pretoria. In the last two decades the SAFARI-1 research reactor has been successfully utilized for the production of radio-isotopes and the neutron transmutation doping of silicon. At the same time, various developments have been undertaken at the horizontal thermal neutron beam line ports. In fulfilling its statuary mandate to apply radiation technology for scientific purposes, Necsa is constantly exploring opportunities to employ the neutrons from its beam line facilities to benefit both academia and industry in research and technological development. This paper outlines the facilities available at SAFARI-1, the current initiatives to establish state-of-the-art user facilities and their application to various fields of material research. (author)

  1. Revamping of control and instrumentation and commissioning and operating experience of pneumatic carrier facility of Dhruva Reactor

    International Nuclear Information System (INIS)

    Dhruva is a 100 MWth research reactor, located at BARC Trombay. It has different facilities for medium and long-term irradiation of samples for research, medical and industrial applications. Self serve and tray rod facilities are for medium term and long-term irradiation of samples ranging from few hours to days or years. However for short-term irradiation of samples of the order of few seconds to few minutes Pneumatic Carrier Facility (PCF) is used. In PCF the capsule (carrying sample) is sent into reactor core and retrieved back by pneumatic force to the experiment room, after elapse of required irradiation time. This paper describes the revamping of C and I systems, design, commissioning and operating experiences and different modifications incorporated based on safety committee recommendations and operating feed back. (author)

  2. Decommissioning of TRIGA-reactors and other nuclear research facilities. Aims of the German working group for incorporation monitoring

    International Nuclear Information System (INIS)

    In 2001 the working group decommissioning was founded in the department incorporation of the German/Switzerland association for radiation protection. Members of the group are representatives of dismantling projects, dismantling companies, organizations of experts, authorities, institutions of measuring incorporations, research facilities and universities. Aims of the group are to guarantee a continuous information and experience exchange between the different decommissioning projects, to establish criteria for facility specific advanced training and further education, to propose requirements for in-vitro incorporation measuring facilities and to draw up a recommendation for incorporation protection of decommissioning projects. It is tried to harmonize the procedure in principle regarding the specific conditions of the different nuclear facilities. The bases for the recommendation are formed by the experiences of the previous dismantling projects in Germany. As an example in this contribution the procedure at the TRIGA reactor of MHH is presented and compared to a facility with alpha-emitters. (author)

  3. Nuclear Safety Research Reactor (NSRR) as a facility for reactor safety research and its modification for the future test plan

    International Nuclear Information System (INIS)

    The NSRR is a modified TRIGA-ACPR (annular core pulse reactor), and attained the initial criticality in May, 1975. It was built for studying reactor fuel behavior under a reactivity-initiated accident condition. The reactor is installed in a pool of 3.6 m width, 4.5 m length and 9 m depth, and water above the reactor core serves as a radiation shield. The reactor core contains 149 driver fuel rods, 6 regulating rods, 2 safety rods and 3 transient rods. An arbitrary reactivity up to 4.67 $ can be set up almost instantaneously in the reactor core. The pulse power generation is terminated by the large negative reactivity induced by prompt temperature feedback without inserting the control rods. This is brought about by an excellent property of the driver fuel which contains 12 wt.% U-ZrH enriched to 20 wt.% U-235. As a unique feature, the NSRR is equipped with a big experimental cavity through the center of the reactor core. It has the diameter of 220 mm, and is called loading tube. It is branched into a vertical loading tube and an offset loading tube. The characteristics of the pulse operation in the NSRR, the outline of fuel irradiation experiment, the future test plan and the modification of the NSRR are described. (Kako, I.)

  4. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  5. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    International Nuclear Information System (INIS)

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy's resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  6. A Versatile Cobalt-60 Irradiation Facility within a Swimming-Pool Research Reactor

    International Nuclear Information System (INIS)

    The IRR-1 enriched-fuel swimming-pool-type reactor incorporates a concrete-shielded gamma cell for using the radiation from spent fuel elements. A versatile 60Co irradiation facility was added at relatively low cost. The 30 000 Ci 60Co source runs on a carriage at the bottom of the reactor pool. The source plaque completely covers the aluminium window between the pool and the irradiation cell. This geometry allows for a ''source-to-target'' overlap; therefore, dose-rate homogeneity within ±20% is attained inside two commercial cases (43 x 31 x 31 cm) or two commercial sacks (80 x 45 x 45 cm) and within ±10% in two flat boxes (40 x 20 x 5 cm) irradiated simultaneously. A set of steel and aluminium screens attached to two rotating turn-tables permits irradiation of commercial cases at any desired dose-rate smaller than 100 000 R/h without the need for turning over at half-time. Two special underwater canisters allow long-term irradiation of flat specimens at dose-rates of less than 600 000 R/h, while the source is used for normal short-term irradiations in the gamma cell. Safety is ensured by a visible and audible indicator and alarm system and by an elaborate interlock system. A system of ionizing gauges and recorders permits measurement of dose-rates over the range 0. 001 to 1 000 000 R/h. Isodose curves for the irradiation chamber have been determined. The cell is soon to be modified to include a refrigeration plant and a timing system for automatic control of source movement. The disadvantages of low source utilization inherent in required source-to-target overlap and of one sided utilization of the radiation are more than compensated for by the possibility of pilot-scale irradiation of commercial cases and by the greater versatility and low cost. This installation is therefore recommended for all similar swimming-pool reactors. It is especially valuable for countries desiring to embark on a food irradiation programme at minimum cost but with maximum

  7. Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

    International Nuclear Information System (INIS)

    This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

  8. A Bayesian approach to unanticipated events frequency estimation in the decision making context of a nuclear research reactor facility

    International Nuclear Information System (INIS)

    Highlights: • The Bayes’ theorem is employed to support the decision making process in a research reactor. • The intention is to calculate parameters related to unanticipated occurrence of events. • Frequency, posterior distribution and confidence limits are calculated. • The approach is demonstrated using two real-world numerical examples. • The approach can be used even if no failures have been observed. - Abstract: Research reactors are considered as multi-tasking environments having the multiple roles of commercial, research and training facilities. Yet, reactor managers have to make decisions, frequently with high economic impact, based on little available knowledge. A systematic approach employing the Bayes’ theorem is proposed to support the decision making process in a research reactor environment. This approach is characterized by low level complexity, appropriate for research reactor facilities. The methodology is demonstrated through the study of two characteristic events that lead to unanticipated system shutdown, namely the de-energization of the control rod magnet and the flapper valve opening. The results obtained demonstrate the suitability of the Bayesian approach in the decision making context when unanticipated events are considered

  9. Operational experiences in radiation protection in fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    The Compact Reprocessing facility for Advanced fuels in Lead cells (CORAL), situated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam is a pilot plant to reprocess the mixed carbide fuel, for the first time in the world. Reprocessing of fuel with varying burn-ups up to 155 G Wd/t, irradiated at Fast Breeder Test Reactor (FBTR), has been successfully carried out at CORAL. Providing radiological surveillance in a fuel reprocessing facility itself is a challenging task, considering the dynamic status of the sources and the proximity of the operator with the radioactive material and it is more so in a fast reactor fuel reprocessing facility due to handling of higher burn-up fuels associated with radiation fields and elevated levels of fissile material content from the point of view of criticality hazard. A very detailed radiation protection program is in place at CORAL. This includes, among others, monitoring the release of 85Kr and other fission products and actinides, if any, through stack on a continuous basis to comply with the regulatory limits and management of disposal of different types of radioactive wastes. Providing radiological surveillance during the operations such as fuel transport, chopping and dissolution and extraction cycle was without any major difficulty, as these were carried out in well-shielded and high integrity lead cells. Enforcement of exposure control assumes more importance during the analysis of process samples and re-conversion operations due to the presence of fission product impurities and also since the operations were done in glove boxes and fume hoods. Although the radiation fields encountered in process area were marginally higher, due to the enforcement of strict administrative controls, the annual exposure to the radiation workers was well within the regulatory limit. As the facility is being used as test bed for validation of prototype equipment, periodic inspection and maintenance of components such as centrifuge

  10. Operational experiences in radiation protection in fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    The COmpact Reprocessing facility for Advanced fuels in Lead cells (CORAL), situated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam is a pilot plant to reprocess the mixed carbide fuel, for the first time in the world. Reprocessing of fuel with varying burn-ups upto 155 GWd/t, irradiated at Fast Breeder Test Reactor (FBTR), has been successfully carried out at CORAL. Providing radiological surveillance in a fuel reprocessing facility itself is a challenging task, considering the dynamic status of the sources and the proximity of the operator with the radioactive material and it is more so in a fast reactor fuel reprocessing facility due to handling of higher burn-up fuels associated with radiation fields and elevated levels of fissile material content from the point of view of criticality hazard. A very detailed radiation protection program is in place at CORAL. This includes, among others, monitoring the release of 85Kr and other fission products and actinides, if any, through stack on a continuous basis to comply with the regulatory limits and management of disposal of different types of radioactive wastes. Providing radiological surveillance during the operations such as fuel transport, chopping and dissolution and extraction cycle was without any major difficulty, as these were carried out in well-shielded and high integrity lead cells. Enforcement of exposure control assumes more importance during the analysis of process samples and re-conversion operations due to the presence of fission product impurities and also since the operations were done in glove boxes and fume hoods. Although the radiation fields encountered in process area were marginally higher, due to the enforcement of strict administrative controls, the annual exposure to the radiation workers was well within the regulatory limit. As the facility is being used as test bed for validation of prototype equipment, periodic inspection and maintenance of components such as centrifuge

  11. A state-of-the-art report on the study of the nuclear reactor thermal hydraulics using integral test facilities

    International Nuclear Information System (INIS)

    Since the integral reactor (SMART) currently under development by KAERI includes distinct design features which are different from those of the conventional large scale commercial reactors, it is necessary to perform integral effect test which will be used to observe overall thermal hydraulic behavior and to verify the safety of the SMART. The integral effect test for the SMART currently promoting by the thermal hydraulic safety research team will provide experimental data to support the reactor design by the performance verification test of the reactor and safety systems, and will provide data to guarantee the safety of SMART design and to verify safety analysis codes for SMART by the integral tests. A proper scaling methodology should be applied to reflect the distinct concepts of the SMART and important physical phenomena should be preserved in this integral test facility. Thus, this report compares the conventional scaling methods and their limitation in detail, and identifies scaling distortions produced practically and examines the methods to remove or minimize the distortion. Also, by comparing in detail the design data of the conventional integral test facilities, this report summarizes comprehensively the limitations, scaling distortions and counter-measures to decrease the distortion. This report is thought to be very useful for the design and manufacturing of the integral test facility for the SMART, and is expected to be used as a guide at the conceptual design and scientific design stages of the integral test facility to simulate the operational and accidental transients to be occurred in the SMART reactor. (author). 44 refs., 27 tabs., 28 figs

  12. The Scottish Research Reactor Centre and its Facilities for the Production and Exploitation of Short-Lived Radioisotopes

    International Nuclear Information System (INIS)

    The Scottish Research Reactor Centre is now under construction and will be completed and in operation by the summer of 1963. The reactor is 100 kW of the tank type with water cooling and water/graphite moderation using enriched U235 fuel. The experimental facilities include a large thermal column, a large shield experiment water tank and a radioisotope production facility with transfer rabbit tubes. There are effectively three through tubes in the central core; one through tube in the thermal column, several small central vertical stringers and one six-inch square vertical stringer penetrating to the centre of the core. Many horizontal stringers pass through the thermal column, the central one penetrating to within one inch of a fuel tank. The reactor facilities are supported by a wide variety of adjacent small laboratories. These include hot source handling and preparation facilities, changing rooms, electrical and mechanical workshops, darkrooms, microcurie laboratories, animal house, biological and chemical laboratories, low-background counting room, lecture theatre and a library. It is expected that the research will extend over many scientific and technological disciplines; a good proportion of the work will involve short-lived radioisotopes and typical projects are described. (author)

  13. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    Science.gov (United States)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  14. Application of high efficiency metal fiber filters in ventilation systems of non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Sand filters, Deep Bed Glass Fiber filters, and remotely replaceable High Efficiency Particulate Air filters have been successfully used for filtration of exhaust air from highly contaminated exhaust air streams. However, none of these technologies satisfy all requirements of an optimum filtration system design. The basic requirements of a nuclear filtration system are a high decontamination factor, low pressure drop, long operating life, sturdiness during normal operation, ability to withstand Design Basis Accidents, minimize generation of waste, minimum maintenance, high radiation resistance, ease of decontamination and decommissioning, and low life cycle cost. High Efficiency Metal Fiber filters are a new technology and provide a suitable alternative to the currently used nuclear air filtration technologies. This article investigates the advantages and disadvantages of the current air filtration technologies and compares them with those of the High Efficiency Metal Fiber filters. High Efficiency Metal Fiber Filters system design considerations for non-reactor nuclear facilities are also discussed in this article. The design considerations include, but are not limited to, physical configuration, space requirements, pressure drop, decontamination factors, dust holding capacity, in-place cleanability, cleaning procedures, in-place testing, and other support equipment. 2 refs., 4 figs

  15. Conflicting opinions: The controversy accompanying the site selection for a German reactor facility (1950-1955)

    International Nuclear Information System (INIS)

    The foundation and history of the Karlsruhe Nuclear Research Center (KfK) partly reflects the history of nuclear energy and the historical development of the German technology sector which in particular with respect to its topical implications (e.g. the acceptance of advanced technologies) is unfortunately still being underestimated in most cases. Studying the historical development of an institution such as Kfk reveals regional ties, municipal, district, neighbourhood and industrial ties which are normally buried and veiled. The history of its foundation and development bears evidence of the fact that Kfk is an integral part of the Karlsruhe technology region. The period from 1952 to 1955 was characterized by vehement disputes preceding the final site selection for the first nuclear reactor facility of the Federal Republic of Germany. Each endeavouring to be selected th competitors, i.e. the Free State of Bavaria represented by the city of Munich and the newly established Southwestern Baden-Wuerttemberg represented by the city of Karlsruhe virtually vied with one another for the most attractive proposals and bids. They were both fully aware of the fact that a positive decision would mean to assume obligations in the order of magnitude of enormous million sums. The slightest chance of being chosen the very region to house the promising and hopeful 'nuclear energy industry' of the Federal Republic of Germany evidently required rather high bids on both sides. (orig./HP)

  16. Neutron spectrum measurements in the aluminum oxide filtered beam facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Neutron spectrum measurements were performed on the aluminum oxide filter installed in the Brookhaven Medical Research Reactor (BMRR). For these measurements, activation foils were irradiated at the exit port of the beam facility. A technique based on dominant resonances in selected activation reactions was used to measure the epithermal neutron spectrum. The fast and intermediate-energy ranges of the neutron spectrum were measured by threshold reactions and 10B-shielded 235U fission reactions. Neutron spectral data were derived from the activation data by two approaches: (1) a short analysis which yields neutron flux values at the energies of the dominant or primary resonances in the epithermal activation reactions and integral flux data for neutrons above corresponding threshold or pseudo-threshold energies, and (2) the longer analysis which utilized all the activation data in a full-spectrum, unfolding process using the FERRET spectrum adjustment code. This paper gives a brief description of the measurement techniques, analysis methods, and the results obtained

  17. Disposal Of Irradiated Cadmium Control Rods From The Plumbrook Reactor Facility

    International Nuclear Information System (INIS)

    Innovative mixed waste disposition from NASA's Plum Brook Reactor Facility was accomplished without costly repackaging. Irradiated characteristic hardware with contact dose rates as high as 8 Sv/hr was packaged in a HDPE overpack and stored in a Secure Environmental Container during earlier decommissioning efforts, awaiting identification of a suitable pathway. WMG obtained regulatory concurrence that the existing overpack would serve as the macro-encapsulant per 40CFR268.45 Table 1.C. The overpack vent was disabled and the overpack was placed in a stainless steel liner to satisfy overburden slumping requirements. The liner was sealed and placed in shielded shoring for transport to the disposal site in a US DOT Type A cask. Disposition via this innovative method avoided cost, risk, and dose associated with repackaging the high dose irradiated characteristic hardware. In conclusion: WMG accomplished what others said could not be done. Large D and D contractors advised NASA that the cadmium control rods could only be shipped to the proposed Yucca mountain repository. NASA management challenged MOTA to find a more realistic alternative. NASA and MOTA turned to WMG to develop a methodology to disposition the 'hot and nasty' waste that presumably had no path forward. Although WMG lead a team that accomplished the 'impossible', the project could not have been completed with out the patient, supportive management by DOE-EM, NASA, and MOTA. (authors)

  18. Simulation of Collimator for Neutron Imaging Facility of TRIGA MARK II PUSPATI Reactor

    Science.gov (United States)

    Zin, Muhammad Rawi Mohamed; Jamro, Rafhayudi; Yazid, Khairiah; Hussain, Hishamuddin; Yazid, Hafizal; Ahmad, Megat Harun Al Rashid Megat; Azman, Azraf; Mohamad, Glam Hadzir Patai; Hamzah, Nai'im Syaugi; Abu, Mohamad Puad

    Neutron Radiography facility in TRIGA MARK II PUSPATI reactor is being upgraded to obtain better image resolution as well as reducing exposure time. Collimator and exposure room are the main components have been designed for fabrication. This article focuses on the simulation part that was carried out to obtain the profile of collimated neutron beam by utilizing the neutron transport protocol code in the Monte Carlo N-Particle (MCNP) software. Particular interest is in the selection of materials for inlet section of the collimator. Results from the simulation indicates that a combination of Bismuth and Sapphire, each of which has 5.0 cm length that can significantly filter both the gamma radiation and the fast neutrons. An aperture made of Cadmium with 1.0 cm opening diameter provides thermal neutron flux about 1.8 x108 ncm-2s-1 at the inlet, but reduces to 2.7 x106 ncm-2s-1 at the sample plane. Still the flux obtained is expected to reduces exposure time as well as gaining better image resolution.

  19. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    International Nuclear Information System (INIS)

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation

  20. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  1. Measuring progress in reactor conversion and HEU minimization towards 2020 - the case of HEU-fuelled research facilities

    International Nuclear Information System (INIS)

    This paper analyzes how to measure progress in the minimization of HEU-fueled research reactors with respect to the International Fuel Cycle Evaluation (INFCE) completed in 1978, and the establishment of new objectives towards 2020. All HEU-fueled research facilities converted, commissioned or decommissioned after 1978, in total more than 310 facilities, are included. More than 130 HEU-fuelled facilities still remain in operation today. The most important measure has been facility shut-down, accounting for 62% of the reduction in U-235 consumption from 1978 to 2007. Presently, only three regions worldwide use significant amounts of HEU; North-America, Russia with the Newly Independent States, and Europe. Projected HEU consumption in 2020 will drop to less 50 kg as the current HEU-fueled steady-state reactors are shut-down or converted. However. if the current lack of concern for HEU in life-time cores is not changed, in particular in Russia, 50-100 such facilities may continue to be in operation in 2020. (author)

  2. Measuring progress in reactor conversion and HEU minimization towards 2020 - The case of HEU-fuelled research facilities

    International Nuclear Information System (INIS)

    The primary impediment that prevents nuclear proliferation is the lack of access to fissile materials. Thus, a recognized objective internationally has been to minimize the use of HEU and reduce the number of locations with HEU present. Yet, nearing the 30 year anniversary of this objective, the number of HEU-fuelled research facilities in operation remains high, HEU is still being used in large quantities, and significant quantities of HEU is still to be found in a large number of unsecured locations worldwide. This paper identifies the most important indicators for measuring progress for the historical and future national and international efforts for research reactor conversion and decommissioning of vulnerable facilities

  3. Characteristics of neutron irradiation facility and dose estimation method for neutron capture therapy at Kyoto University research reactor institute

    International Nuclear Information System (INIS)

    The neutron irradiation characteristics of the Heavy Water Neutron Irradiation Facility (HWNIF) at the Kyoto University Research Reactor Institute (KIJRRI) for boron neutron capture therapy (BNCT), is described. The present method of dose measurement and its evaluation at the KURRI, is explained. Especially, the special feature and noticeable matters were expounded for the BNCT with craniotomy, which has been applied at present only in Japan. (author)

  4. Characteristics of UO2-Zircaloy fuel rod materials from the Saxton reactor for use in power burst facility

    International Nuclear Information System (INIS)

    This report describes the destructive and nondestructive characterization studies performed on Zircaloy-4-clad UO2 fuel rods taken from the Saxton reactor. The results of the studies and a description of the techniques used in performing the studies are included. The fuel rods will now be used in a series of experiments to be conducted in the Power Burst Facility at the Idaho National Engineering Laboratory

  5. Characteristics of UO/sub 2/-Zircaloy fuel rod materials from the Saxton reactor for use in power burst facility

    Energy Technology Data Exchange (ETDEWEB)

    Gibson, G.W.; Murdock, B.A.; Quapp, W.J.; Hobbins, R.B.; Ising, R.H. (comps.)

    1976-09-01

    This report describes the destructive and nondestructive characterization studies performed on Zircaloy-4-clad UO/sub 2/ fuel rods taken from the Saxton reactor. The results of the studies and a description of the techniques used in performing the studies are included. The fuel rods will now be used in a series of experiments to be conducted in the Power Burst Facility at the Idaho National Engineering Laboratory.

  6. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  7. Model experiments on simulation of the WWER water-chemical conditions at loop facilities of the MIR reactor

    International Nuclear Information System (INIS)

    The experiments on simulation of the WWER type reactors water-chemical conditions have been started at the State Scientific Center RIAR. These experiments are being conducted at the multi-loop research MIR reactor at the PVK-2 loop facility. The dosage stand was created. It allows introduction of boric acid, potassium and lithium hydroxides, ammonia solutions and gaseous hydrogen. Corrosion tests of the Russian E-635 and E-110 alloys are being conducted at the PVK-2 loop under the WWER water-chemical conditions. If necessary, fuel elements are periodically extracted from the reactor to perform visual examination, to measure their length, diameter, to remove the deposits from the claddings, to measure the burnup and to distribute the fission products over the fuel element by gamma-spectrometry. The chemical analytical 'on line' equipment produced by the ORBISPHERE Laboratory (Switzerland) will be commissioned in the nearest future to measure concentration of the dissolved hydrogen and oxygen as well as pH and specific conductivity. The objective of the report is to familiarize the participants of the IAEA Technical Committee with the capabilities of performing the model water-chemical experiments under the MIR reactor loop facility conditions. (author)

  8. The Fission Converter-Based Epithermal Neutron Irradiation Facility at the Massachusetts Institute of Technology Reactor

    International Nuclear Information System (INIS)

    A new type of epithermal neutron irradiation facility for use in neutron capture therapy has been designed, constructed, and put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). A fission converter, using plate-type fuel and driven by the MITR, is used as the source of neutrons. After partial moderation and filtration of the fission neutrons, a high-intensity forward directed beam is available with epithermal neutron flux [approximately equal to]1010 n/cm2.s, 1 eV ≤ E ≤ 10 keV, at the entrance to the medical irradiation room, and epithermal neutron flux = 3 to 5 x 109 n/cm2.s at the end of the patient collimator. This is currently the highest-intensity epithermal neutron beam. Furthermore, the system is designed and licensed to operate at three times higher power and flux should this be desired. Beam contamination from unwanted fast neutrons and gamma rays in the aluminum, polytetrafluoroethylene, cadmium and lead-filtered beam is negligible with a specific fast neutron and gamma dose, Dγ,fn/φepi [less than or approximately equal] 2 x 10-13 Gy cm2/nepi. With a currently approved neutron capture compound, boronophenylalanine, the therapeutically advantageous depth of penetration is >9 cm for a unilateral beam placement. Single fraction irradiations to tolerance can be completed in 5 to 10 min. An irradiation control system based on beam monitors and redundant, high-reliability programmable logic controllers is used to control the three beam shutters and to ensure that the prescribed neutron fluence is accurately delivered to the patient. A patient collimator with variable beam sizes facilitates patient irradiations in any desired orientation. A shielded medical room with a large window provides direct viewing of the patient, as well as remote viewing by television. Rapid access through a shielded and automatically operated door is provided. The D2O cooling system for the fuel has been conservatively designed with excess

  9. Neutron depolarization facility at the research reactor IRT-Sofia. Study of amorphous ferromagnetic ribbons

    International Nuclear Information System (INIS)

    The design of the setup for neutron depolarization studies and results of its test on the research reactor IRT-Sofia are briefly described. The characteristics of this facility permit it to be used effectively in physical experiments. As an example the results of the study of amorphous ferromagnetic ribbons are presented in order to demonstrate the sensitivity of the technique. The samples used are of magnetic amorphous alloys in the form of ribbons fabricated by means of rapid quenching of the melt ejected upon a water cooled copper drum rotating at a frequency 1 kHz. A set of ribbon pieces of the amorphous alloy Fe48.5Ni34P17.5 stacked in a container made by copper strips are used as a reference sample. In all measurements the incident polarized neutron beam is perpendicular to the ribbon plane so that the polarization vector lies in the plane. The results are analyzed by means of the Maleyev-Ruban's formula which gives the relationship between the initial polarization and the final polarization of the neutron beam passing through the sample. The dependence of the depolarization of the polarized neutron beam transmitted through the reference sample on the sample thickness is presented. The depolarization of samples of different compositions as a function of thickness is shown. Spectral analysis of the polarization of the neutron beam transmitted through the sample magnetized by an external field B=100 Oe which is applied perpendicular to the incident polarization vector and the neutron velocity is carried out. It is concluded that the 'intrinsic' depolarization is constant and wavelength independent. 11 figs., 1 tab., 13 refs. (orig.)

  10. Design of fuel fabrication plant of Fast Reactor Fuel Cycle Facility for reload requirement of Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    India's economic growth is on a fast growth track. The energy demand is expected to grow rapidly in the coming decades. The growth in population and economy is creating huge demand for energy which has to be met with environmentally benign technologies. Nuclear energy is best suited to meet this demand in a sustainable manner without causing undue environmental impact. Fast reactors are expected to be major contributors in sufficing this demand to a great extent. As an effort to achieve the objective, a Prototype Fast Breeder Reactor is being constructed at Kalpakkam. This paper also highlights the design features of FFP, unit operations, scheme of automation, branched layout of glove box train, shielding arrangement on glove boxes, accident consequence analysis etc.

  11. The Text of the Agreement Between the Agency and Argentina for the Application of Safeguards to the Embalse Power Reactor Facility

    International Nuclear Information System (INIS)

    The text of the Agreement between the Agency and the Government of the Republic of Argentina for the Application of Safeguards to the Embalse Power Reactor Facility is reproduced in this document for the information of all Members

  12. 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)

    The text of the Agreement between the Agency and the Government of the Republic of Argentina for the Application of Safeguards to the Atucha Power Reactor Facility is reproduced in this document for the information of all Members

  13. Incorporating the operation of a small research reactor facility to support a national nuclear power program

    International Nuclear Information System (INIS)

    A small research reactor similar to the Penn State Breazeale Nuclear Reactor, which is a 1 Megawatt TRIGA, can be invaluable in supporting a national nuclear power programme. The research reactor provides an operating reactor for training nuclear engineers, nuclear operators, and other nuclear specialists required to construct, operate, and maintain the nuclear power plant. When operation of the power plant begins, highly trained, well-qualified and competent personnel will be available to operate, supervise, and maintain a safe and efficient power plant operation. The paper describes the organization of a nuclear science centre as well as research reactor activities. An example of an actual two week training programme of an electric utility is included with a list of experiments and demonstrations. University projects and experiments using the reactor are listed. (author)

  14. The Oarai Branch of IMR, Tohoku University as open facility for university researchers utilizing fission reactors

    International Nuclear Information System (INIS)

    For advanced future research activities utilizing fission reactors and hot laboratories, effective interlinks among fission reactors and hot laboratories are indispensable. Oarai Branch of Institute for Materials Research in Tohoku University has been playing an important role for supplying related tools for university researchers, in fission reactor irradiation and post irradiation examinations, under tight collaboration with JAERI and JNC. Now the Oarai Branch is planning to expand its collaborative functions, utilizing multi-reactors over the world and making effective interlinks among related hot laboratories in several institutions. The talk will give rough view of the present plan of the Oarai Branch, IMR, Tohoku University for tight and effective collaboration among institutions. (author)

  15. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    International Nuclear Information System (INIS)

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants

  16. Neutronic temperature determination in irradiation facilities of the RP-0 nuclear reactor

    International Nuclear Information System (INIS)

    RP-10 a swimming pool MTR type research reactor that attained power of 10 MW. The neutron temperature has been measurement in four different positions of the reactor at 300Kw of power, for this purpose was used the direct method, proposed for De Corte et al. The results were presented and discussed. (authors)

  17. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

    2013-09-15

    Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

  18. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  19. Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, first half of fiscal year 1981

    International Nuclear Information System (INIS)

    The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 57 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as neutron radiography using a research reactor, measurement of Zr/Hf ratio in zirconium, interstitial germanium atoms in thermal neutron irradiation study, measurement of induced radioactivity due to neutrons in Nagasaki and Hiroshima atomic bombings, properties of semiconductor electrons in radiation study, induction of mutation in crops by neutron irradiation and utilization for breeding, thermal fluorescence mechanism of alkali halide and MgO single crystals, atomic configuration in PZT rhombohedron phase, modulated structure of Cu-Co alloys, excitation of nuclei by positron annihilation and others. (Kako, I.)

  20. Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

    International Nuclear Information System (INIS)

    A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

  1. Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1981

    International Nuclear Information System (INIS)

    The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 61 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as radioactivation analysis of trace elements in rocks and minerals, anodic oxidation films of GaAs and structure, measurement of yield of uranium isotopes produced by reactor neutron irradiation of thorium, geochemical study of trace elements in hydrosphere by radio-activation analysis, various diseases and variation of elements in rat furs, Moessbauer spectroscopic study of gold compounds with singular coupling by Au-197, measurement of grass-eating quantity and rate of digestion of cows using Au and Eu, sickness biochemical study of trace elements in hair samples of patients and others. (Kako, I.)

  2. Renovation of CPF (Chemical Processing Facility) for Development of Advanced Fast Reactor Fuel Cycle System

    International Nuclear Information System (INIS)

    CPF (Chemical Processing Facility) was constructed at Nuclear Fuel Cycle Engineering Laboratories of JAEA (Japan Atomic Energy Agency) in 1980 as a basic research field where spent fuel pins from fast reactor (FR) and high level liquid waste can be dealt with. The renovation consists of remodeling of the CA-3 cell and the laboratory A, installation of globe boxes, hoods and analytical equipments to the laboratory C and the analytical laboratory. Also maintenance equipments in the CA-5 cell which had been out of order were repaired. The CA-3 cell is the main cell in which important equipments such as a dissolver, a clarifier and extractors are installed for carrying out the hot test using the irradiated FR fuel. Since the CPF had specialized originally in the research function for the Purex process, it was desired to execute the research and development of such new, various reprocessing processes. Formerly, equipments were arranged in wide space and connected with not only each other but also with utility supply system mainly by fixed stainless steel pipes. It caused shortage of operation space in flexibility for basic experimental study. Old equipments in the CA-3 cell including vessels and pipes were removed after successful decontamination, and new equipments were installed conformably to the new design. For the purpose of easy installation and rearranging the experimental equipments, equipments are basically connected by flexible pipes. Since dissolver is able to be easily replaced, various dissolution experiments is conducted. Insoluble residue generated by dissolution of spent fuel is clarified by centrifugal. This small apparatus is effective to space-saving. Mini mixer settlers or centrifugal contactors are put on to the prescribed limited space in front of the backside wall. Fresh reagents such as solvent, scrubbing and stripping solution are continuously fed from the laboratory A to the extractor by the reagent supply system with semi-automatic observation

  3. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

  4. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  5. Studies and research concerning BNFP: life of project operating expenses for away-from-reactor (AFR) spent fuel storage facility. Final report

    International Nuclear Information System (INIS)

    Life of Project operating expenses for a licensed Away-From-Reactor (AFR) Spent Fuel Storage Facility are developed in this report. A comprehensive business management structure is established and the functions and responsibilities for the facility organization are described. Contractual provisions for spent fuel storage services are evaluated

  6. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  7. Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities

    International Nuclear Information System (INIS)

    A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the core in good agreement with calculations performed using the deterministic code CITATION and measurements using activation foils. The model is used for the prediction of the neutron field characteristics at the reactor irradiation devices and enables the design and evaluation of experiments involving material irradiations

  8. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  9. Testing of the Y-12 Plant Criticality Accident Alarm System detectors at the Sandia Pulsed Reactor Facility

    International Nuclear Information System (INIS)

    The Oak Ridge Y-12 Plant operates its Criticality Accident Alarm System (CAAS) according to the guidance of Standard ANSI/ANS-8.3-1986. This standard requires that the detector shall not fail to initiate an alarm when subjected to a radiation field of at least 0.1 Gy/s (10 rad/s). It also requires that the system shall be designed to immediately detect the minimum accident of concern and shall produce an alarm within one half second of activation. Sixty-three new detectors that use plastic scintillators have been obtained to upgrade the current Y-12 Plant CAAS. To ensure that these detectors can support the above criteria, testing was done using the SPR III reactor at the Sandia Pulsed Reactor Facility

  10. International cooperation in converting the Philippine Research Reactor PRR-1 and the impacts of its upgraded facilities

    International Nuclear Information System (INIS)

    The paper describes the institutional and physical considerations in the fuel conversion and upgrade in power of the Philippine Research Reactor PRR-1. Various purposes for the conversion and upgrade discussed in the paper can be summarized as production of short-lived isotopes, neutron activation analysis enhancement for an in-house research and analytical service to customers, delayed neutron activation analysis of geological samples from uranium exploration, enhancement of beam tube fluxes for physics experiments, and training. The fuel conversion was from a highly enriched uranium (HEU, 93%), aluminum clad MTR plate type fuel to low enriched (LEU, 20%). Incoloy 800 clad TRIGA uranium-zirconium hydride (UZrH) four-rod fuel clusters. The UZrH clusters were designed to directly replace the plate elements in the original grid plate. Thus no core structure modifications were required. The upgrade in power was from 1 MW to 3 MW. This required an increase in coolant pumping power and flow and an increase in the capacity of the heat exchanger and cooling tower system to reject the additional heat load. The various reactor instrumentation systems were also replaced with an integrated system for reactor monitoring and control. International assistance and cooperation from other countries and the International Atomic Energy Agency have aided the overall facility operation and expertise and the conversion and upgrade through funding grants, expert help and fellowships. The new reactor capabilities include in-core experimental facilities with thermal flux values nearly six times greater than that of the original core, pulsing operation, and increased performance and enhanced safety generally. (author). 3 figs, 1 tab

  11. A nuclear desalination complex with a VK-300 boiling type reactor facility

    International Nuclear Information System (INIS)

    RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m3/day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

  12. Identification of a leaking TRIGA fuel element at the reactor facility of Pavia

    International Nuclear Information System (INIS)

    On January 28th 2004, during a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark II reactor of the University of Pavia a small but detectable amount of 137Cs contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought at the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analyzed in a low background gamma-ray detector. As a result a small amount of fission products were detected in the reactor pool water (few Bq/g) suggesting the existence of a possible clad defect in one ore more fuel elements. As a consequence of this situation a campaign of gamma-ray spectrometry was implemented in order to evaluate the importance of the release. Analyzes using a HGe detector (1.72 keV FWHM - 31.3 % efficiency - 58.5 Photo Peak/Compton) were performed and the most significant results are presented as well as the identification of the leaking fuel element. The fission products leakage was due to a micro-fissure of a fuel element that released only noble gas when it was heated up to a temperature around 90oC , i.e. at the reactor power of about 100 kW. The oldest SST clad instrumented fuel element in the core was identified as the origin of the release. It was removed from its position and stored in a rack of the reactor pool under 4 m of water shield. The reactor came back in regular operation on March 22nd 2004 and no other fission products leakages were detected. After this situation the reactor pool water is sampled and measured with a low-background gamma-ray detector every month before the reactor start-up and after one hour of operation of the reactor at full nominal power. (nevyjel)

  13. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    Energy Technology Data Exchange (ETDEWEB)

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

  14. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  15. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  16. HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommission Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site

    International Nuclear Information System (INIS)

    This paper contains a summary of the holdup and material control and accountability assays conducted for the determination of highly enriched uranium in the deactivation and decommissioning of Building 321 -M at the Savannah River Site. The facility was the Reactor Fuel Fabrication Facility at SRS and assemblies and miscellaneous components for the SRS production reactors. The facility operated for 25 years. During this time thousands of uranium-aluminum-alloy production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The D and D project is likely to represent an important example for D and D activities across SRS and across the Department of Energy weapons complex

  17. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw; K. G. Condie; S. Curtis Wilkins

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  18. The Prompt Gamma Neutron Activation Analysis Facility at the RA-6 reactor of the Bariloche Atomic Centre, Argentina

    International Nuclear Information System (INIS)

    The RA-6 is a research reactor with 500 kW of thermal power, located at the Bariloche Atomic Centre. In one of its five extraction tube facilities a prompt gamma neutron activation analysis system is now under construction. The neutron thermal flux in the position sample is 7 106 n/cm2s using a 5 cm thick bismuth filter. This work presents two facility designs, a preliminary one and another one with some improvements. Shielding optimizing experiences which justify the incorporated improvements are described. The applications of them allow the measurement of a borated sample. Also presented is a new design of the beam catcher and it is compared with the old one by MCNP modelling. New applications are being considered in the frame of the contract with the IAEA under the Co-ordinated Research Project (CRP) on 'New Applications of PGNAA'. (author)

  19. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  20. The 30 kW research reactor facility in Ghana: past, present and future programmes

    International Nuclear Information System (INIS)

    The Ghana Research Reactor-1 (GHARR-1) is a small, simple, reliable and safe reactor design and constructed by China Institute of Atomic Energy (CIAE). GHARR-1 adopts the pool-tank structure and employs highly enriched uranium as fuel, light water as moderator and coolant, metal beryllium as reflectors. The reactor is cooled by natural convention. The rated maximum thermal power of GHARR-1 is 30 kW; the corresponding neutron flux is 1.0x1012 cm-2s-1. The refueling mode of the reactor is to totally change the old core with a new one, the lifetime being more than ten years. Since the commencement of operation of the low-flux miniature neutron source reactor (MNSR) in 1995, a significant number of research and development in the field of neutron activation analysis have taken place. During its 12 years of operation, after the first criticality, the reactor has been used as a neutron source for research, teaching and training to support several graduate and post graduate careers for students from universities in Ghana and the West African sub-region. Owing to the stable flux of the reactor and rapid proliferation in utilization, several analytical techniques have been developed. The GHARR-1 application in neutron activation analysis included: (i) Food analysis; (ii) Heavy metals determination in environmental samples; (iii) Determination of major, minor and trace elements in geological samples; (iv) And mineral prospecting among others. The educational programmes in place at the center are teaching and learning in nuclear engineering, nuclear physics, nuclear and radiochemistry and other related fields. (author)

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

    International Nuclear Information System (INIS)

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

  2. Design of Stopper of Prompt Gamma Neutron Activation Analysis Facility at China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The PGNAA facility consists of the filtered collimated neutron beam, the shielding of the whole facility, the control system, the detecting equipment and the data acquisition and analysis system. The neutron beam is filtered by a mono-crystalline bismuth filter,

  3. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1992. Twenty-fifth annual report, Volume 14

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv)

  4. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1990: Twenty-third annual report

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1990. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1990 annual reports submitted by about 443 licensees indicated that approximately 214,568 individuals were monitored, 110,204 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.19 rem (cSv) and an average measurable dose of about 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,361 individuals completed their employment with one or more of the 443 covered licensees during 1990. Some 77,633 of these individuals terminated from power reactor facilities, and about 11,083 of them were considered to be transient workers who received an average dose of 0.67 rem (cSv)

  5. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1993. Volume 15, Twenty-six annual report

    International Nuclear Information System (INIS)

    This report the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1993. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1993 annual reports submitted by about 360 licensees indicated that approximately 189,711 individuals were monitored, 169,872 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measured dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 99,749 individuals completed their employment with one or more of the 360 covered licensees during 1993. Some 91,000 of these individuals terminated from power reactor facilities, and about 12,685 of them were considered to be transient workers who received an average dose of 0.49 rem (cSv)

  6. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1984. Seventeenth annual report. Volume 6

    International Nuclear Information System (INIS)

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1984. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1984 annual reports submitted by about 500 licensees indicated that approximately 195,000 individuals were monitored, 171,000 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.30 rem (cSv) and an average measureable dose of 0.55 rem (cSv). Termination radiation exposure reports were analyzed to reveal that about 67,500 individuals completed their employment with one or more of the 500 covered licensees during 1984. Some 66,100 of these individuals terminated from power reactor facilities, and about 5500 of them were considered to be transient workers who received an average dose of 0.91 rem

  7. Performance and experience at neutron radiography facility around Pakistan Research Reactor-I

    International Nuclear Information System (INIS)

    Neutron radiography facility at PINSTECH has been used for the inspection of couple of test pieces. Thermal neutron flux level has been measured at the object plane. The neutron to gamma ratio at the object plane is greater than 5x 10/sup 4/ n.cm/sup -2/. mR/sup -1/. Cadmium ratio of 18 has been measured experimentally. The collimating ratio is 60. Various constituents of the neutron beam have been measured by exposing beam purity indicator. To determine the sensitivity of the technique and categories the class of the facility, sensitivity indicator has been fabricated and exposed to the thermal neutron beam. All the gaps in the sensitivity indicator are visible. The characterisation of the facility has helped asses the status of the facility for direct neutron radiography. Some modification in the design of the facility are planned to increase the thermal neutron flux level to reduce the exposure time to couple of minutes. (author)

  8. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  9. ISO-9001: An approach to accreditation for an MTR facility: SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    The SAFARI-1 Research Reactor obtained ISO-9001 accreditation via the South African Bureau of Standards in September 1998. In view of the commercial applications of the reactor, the value of acquisition of the accreditation was considered against the cost of implementation of the Quality System. The criteria identified in the ISO-9001 standard were appraised and a superstructure derived for management of the generation and implementation of a suitable Quality Management System (QMS) for the fairly unique application of a nuclear research reactor. A Quality Policy was established, which formed the basis of the QMS against which the various requirements and/or standards were identified. In addition, since it was considered advantageous to incorporate the management controls of Conventional and Radiological Safety as well as Plant Maintenance and Environmental Management (ISO 14001), these aspects were included in the QMS. (author)

  10. Distribution of 16N and 19O in the reactor pool water of the THOR facility

    International Nuclear Information System (INIS)

    Radioactive 16N and 19O in the Tsing Hua Open-Pool Reactor, produced from 16O(n,p)16N and 18O(n,γ)19O reactions, respectively, have been measured using a rapid sampling device and gamma-ray spectroscopic systems. The radioactivity of the 7-s half-life 16N and 27-s half-life 19O in the pool water are monitored in the power range from 1 W to 1 MW. The three-dimensional concentration of these radionuclides in the water coolant is also contour mapped down to the detection limit of 10 Bq/l. The spatial distribution of the short-lived radionuclides in the reactor pool, resulting from both the neutron flux distribution and heat transfer characteristics external to the core, is discussed for reactor operation at various power levels

  11. Application of a triga research reactor as the neutron source for a production neutron radiography facility

    International Nuclear Information System (INIS)

    GA Technologies Inc. (GA) has developed a Stationary Neutron Radiography System (SNRS) using a 250-1000 KW TRIGA reactor as the neutron source. The partially below ground reactor will be equipped with four vertical beam tubes originating in the reactor graphite reflector and installed tangential to the core to provide a strong current of thermal neutrons with minimum gamma-ray contamination. The vertical beam tubes interface with rugged component positioning systems designed to handle intact F-111 aircraft wings, partial A-10 aircraft wings, pyrotechnics, and other honeycomb aircraft structures. The SNRS will be equipped with real-time, near-real-time, and film-radiographic imaging systems to provide a broad spectrum of capability for detection or corrosion of entrained moisture in large aircraft panels. (author)

  12. Application of a triga research reactor as the neutron source for a production neutron radiography facility

    International Nuclear Information System (INIS)

    GA Technologies Inc. (GA) has developed a Stationary Neutron Radiography System (SNRS) using a 250-1000 kW TRIGA reactor as the neutron source. The partially below ground reactor will be equipped with four vertical beam tubes originating in the reactor graphite reflector and installed tangential to the core to provide a strong current of thermal neutrons with minimum gamma-ray contamination. The vertical beam tubes interface with rugged component positioning systems designed to handle intact F-11 aircraft wings, partial A-10 aircraft wings, pyrotechnics, and other honeycomb aircraft structures. The SNRS will be equipped with real-time, near-real-time, and film-radiographic imaging systems to provide a broad spectrum of capability for detection of corrosion or entrained moisture in large aircraft panels

  13. Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa

    International Nuclear Information System (INIS)

    At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables

  14. Report of researches by common utilization of facilities in Kyoto University Research Reactor Institute, latter half of fiscal year 1982

    International Nuclear Information System (INIS)

    The technical report of the Kyoto University Research Reactor Institute is published any time to immediately report on the results of the functional tests of various experimental facilities, the test results for the products made for trial, radiation control, the situation of waste treatment, the data required for research and experiment such as the reports of study meetings, the conspicuous results obtained amid researches, new processes, and the discussion on other papers and reports. In this report, the title, the names of reporters and the summary of 65 researches carried out by the common utilization of the facilities in the Kyoto University Research Reactor Institute are collected. The themes of the researches are such as Moessbauer spectroscopic study of ferrocene and its derivative iodides by I-129, decomposition of cadmium telluride during heat treatment, element distribution in resource living things and environmental substances produced in northern ocean, radioactivation analysis of trace elements in blood of tumor-bearing animals, radioactivation analysis of noble metal elements in geochemical samples, relaxation phenomena by gamma-gamma perturbation angle correlation, separation of components in Allende meteorite and their radioactivation analysis, measurement of cross section of Pa-231 (n, gamma) reaction and others. (Kako, I.)

  15. ISO 9001 and ISO 14001: An Integrated Quality Management System for an MTR Facility SAFARI-1 Research Reactor

    International Nuclear Information System (INIS)

    The SAFARI-1 research reactor, owned and operated by the South African Nuclear Energy Corporation (Necsa), initially obtained ISO 9001 accreditation of its Quality, Health, Safety and Environmental (QHSE) management system via international affiliation from the South African Bureau of Standards (SABS) during 1998 and re-certification according to ISO 9001 (2000) in 2003. With ever-increasing demands on nuclear facilities to demonstrate conformance to environmental policies, SAFARI-1 has now developed an Environmental Management System (EMS) that is compliant with ISO 14001 (1996) and is fully integrated with the SAFARI-1 Quality Management System (QMS). The dynamic involvement of SAFARI-1 in commercial applications demanded that any transition of the original QMS to a fully incorporated QHSE system had to be done in a way that would ensure sustained delivery of a safe and reliable service with continuous quality. At the same time, the primary vision of operating a facility under an efficient financial management programme was essential. The criteria established by the original ISO 9001 compliant QMS were appraised against the additional requirements of ISO 14001 and a suitable superstructure derived for generation and implementation of an inclusive EMS. The transitional integration of this system was planned so as to produce a QMS suitable to quality, environmental and other management related issues for application to the unique function of a nuclear research reactor. (author)

  16. Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

    International Nuclear Information System (INIS)

    Highlights: ► In a DT fusion machine several events will generate highly tritiated water (HTW). ► PERMCAT appears a promising process to recover tritium from HTW. ► In order to perform R and D activity on HTW processing with PERMCAT, such water has to be produced on purpose. ► A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min−1 of HTW with very high specific tritium activity. ► The paper presents the inactive commissioning of the μCCR required before the integration in CAPER facility. ► The combination of the μCCR with the O2 sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. - Abstract: In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose. Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min−1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 1016 Bq kg−1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water. The combination of the μCCR with the O2 sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R and D activity on

  17. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 108 n/ cm2/ s. According to IAEA (2001) flux of 1.00 x 109 n/ cm2/ s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  18. Technical capabilities of BOR-60 facility and research programme on BN-reactors safety

    International Nuclear Information System (INIS)

    In the scientific and research institute of nuclear reactors (town Dimitrovgrad, Russia) is created research base, consisted from the reactor BOR-60 and the complex of material testing cameras and technological installations of fuel cycle. The works on followed directions are being carried out on this technical and scientific base: the studies of fuel and fuel elements; the study of the problems of fuel cycle; the study of constructed and absorbed materials; the testing of new equipment, improvement safety tools and operation properties. The above mentioned works have been carried out complex character, have been oriented on the decision of current tasks in atomic energetics

  19. Training methods and facilities on reactor and simulators at the Grenoble Nuclear Research Centre

    International Nuclear Information System (INIS)

    Siloette is a CEA unit with a threshold vocation: operation of the Siloette 100 KW pool-type research reactor; basic training in reactor physics for nuclear power plant operators; and production of nuclear power plant simulators: PWR, GCR and more generally of all types of industrial unit simulators, thermal power plant, network, chemical plant, etc. From this experience, they would emphasize in particular the synergy arising from these complementary activities, the essential role of training in basic principles as a complement to operation training, and the ever-increasing importance of design ergonomics of the training means

  20. Evaluation of nuclear facility decommissioning projects. Project summary report, Elk River Reactor

    International Nuclear Information System (INIS)

    This report summarizes information concerning the decommissioning of the Elk River Reactor. Decommissioning data from available documents were input into a computerized data-handling system in a manner that permits specific information to be readily retrieved. The information is in a form that assists the Nuclear Regulatory Commission in its assessment of decommissioning alternatives and ALARA methods for future decommissionings projects. Samples of computer reports are included in the report. Decommissioning of other reactors, including NRC reference decommissioning studies, will be described in similar reports

  1. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  2. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  3. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  4. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor defueling and disassembly. Summary status report. Volume 3

    International Nuclear Information System (INIS)

    This document summarizes information relating to the preparations for defueling the Three Mile Island Unit 2 (TMI-2) reactor and disassembly activities being performed concurrently with decontamination of the facility. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data sysem which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during the period of April 23, 1979 to April 16, 1985, in the completion of activities related to preparation for reactor defueling. Support activities conducted outside of radiation areas are not included within the scope of this report. Computerized reports included in this document are: A chronological summary listing work performed for the period; and summary reports for each major task undertaken in connection with the specific scope of this report. Presented in chronological order for the referenced time period. Manually-assembled table summaries are included for: Labor and exposures by department; and labor and exposures by major activity

  5. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    International Nuclear Information System (INIS)

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of 232Th and 237Np, measured in GFR-like lattices. (authors)

  6. ISO 9001 and ISO 14001: An integrated quality management system for an MTR facility: SAFARI-1 research reactor

    International Nuclear Information System (INIS)

    Full text: The SAFARI-1 Research Reactor, which is owned and operated by NECSA, initially obtained ISO 9001 accreditation of its Quality Management System (QMS) via international affiliation from the South African Bureau of Standards (SABS) during 1998 and re-certification according to ISO 9001 (2000) this year. With ever-increasing demands on nuclear facilities to demonstrate conformance to environmental policies, SAFARI-1 has now developed an Environmental Management System (EMS) that is compliant with ISO 14001 (1996) and is fully integrated with the SAFARI-1 QMS under the controlling implementation of ISO 9001 (2000). The dynamic involvement of SAFARI-1 in commercial applications demanded that any transition of the original QMS had to be done in a way that would ensure sustained delivery of a reliable service with continuous quality. At the same time the primary vision of operating a safe facility under an efficient financial management program was essential. The criteria established by the original ISO-9001 compliant QMS were appraised against the additional requirements of ISO 14001 and a suitable superstructure derived for generation and implementation of an inclusive ISO 14001 EMS. The transitional integration of this system was planned so as to produce a QMS suitable to quality, environmental and other management related issues for application to the fairly unique function of a nuclear research reactor. An amended Quality Policy was established, which with consensus of all concerned, now forms the basis of the integrated QMS against which the various requirements and/or standards were identified. In the case of SAFARI-1 the following requirements were adopted: ISO 9001 (2000): Quality Management Systems; ISO 14001 (1996) Environmental Management Systems; IAEA SS No. 50-C/SG-Q Quality Assurance for Safety in Nuclear Power Plants and Nuclear Installations; NL-27: Nuclear Licence (Incorporating SAFARI-1 at NECSA); RM-PRG-0100 NECSA Health, Safety and

  7. Sodium removing facility for core-constitutional elements of FBR type reactor

    International Nuclear Information System (INIS)

    Reactor core-constitutional elements as spent reactor core fuel assemblies are contained in a containing vessel. An inert gas (N2, Ar or He) is filled in the containing vessel through an inert gas supply channel. The temperature of the inert gas is raised by the remaining after heat of the reactor core-constitutional elements. The inert gas is circulated and heated through a preheating circuit by driving a recycling gas blower and returned to the containing vessel. If the inert gas is heated to a predetermined temperature, metal sodium deposited on the surface of the materials of the reactor core-constitutional elements is evaporated. Next, a vacuum pump unit of a vacuum exhaustion channel is driven to suck an inert gas entraining sodium vapor in the containing vessel, and the sodium vapor is cooled, condensed thereby separated in a sodium separator. Then, the inert gas at a low temperature is introduced to a vacuum exhaustion channel to remove and discharge remained sodium vapor by a sodium trap. (I.N.)

  8. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  9. Physical inventory verification exercise at a light-water reactor facility

    International Nuclear Information System (INIS)

    A simulated physical inventory verification exercise was performed at the Three Mile Island (TMI) Unit 1 reactor. Inspectors from the Internatinal Atomic Energy Agency made measurements on fresh- and spent-fuel assemblies and verified the special nuclear material inventory at TMI. Simulated inspection log sheets and computerized inspection reports were prepared

  10. Design project of the experimental facility for testing uranium creep in the reactor

    International Nuclear Information System (INIS)

    This report contains the design for constructing the experimental device for testing metal uranium creep in the RA reactor core under defined neutron flux conditions, temperature, mechanical loads and time of irradiation. This device will be placed in one of the experimental channels in the core. This report contains physical, thermal and mechanical calculations and engineering drawings of the device

  11. Performance Indicator Program for US Department of Energy reactors and facilities

    International Nuclear Information System (INIS)

    The US Department of Energy (DOE) is developing a Performance Indicator (PI) Program for all facilities. The objective is to periodically collect, statistically analyze and present performance-related information in a concise and consistent format for DOE and safety of facility operations. A set of 14 DOE-Hq. defined PI's has been established after review of programs used by other organizations. Since July 1989, these PI's have been used in a trial program for eight diverse DOE facilities. Electronic reporting is directly to the DOE Safety Performance Measurement System Computer. Results have demonstrated the feasibility and usefulness of a DOE-wide PI Program and steps are being taken to include all DOE facilities. 4 refs., 2 figs., 2 tabs

  12. Research nuclear reactor and particle accelerator as complementary facilities in obtaining experimental nuclear data

    International Nuclear Information System (INIS)

    In the last decade a large amount of diverse and high precision nuclear data is in high demand to support both power applications (in nuclear fusion and fission reactors, fuel cycle in all its stages, nuclear safety) and non-power applications (radiation dosimetry, life sciences, ecology, industry, etc.) of atomic and nuclear techniques. The atomic and nuclear data are generated from experimental measurements, theoretical model calculations and data evaluation, which are finally validated internationally and included in data bases under standardized formats. Measuring of these data imply utilization of research reactors and charged particle accelerators, in complex experiments characterized by high degree of complementarity. Aspects of this complementarity in the nuclear data obtained from reactors and accelerators will be presented in this work. In Romania an advanced research reactor (TRIGA at INR Pitesti) and an electrostatic 4-5 MeV/nucleon accelerator (TANDEM Van de Graaff at IFIN - HH, Bucharest) are operational and a rich scientific expertise in the field of nuclear structure and reaction mechanisms is available. Consequently, the paper considers a project at a national scale for measuring and evaluating nuclear data. Having in view numerous signals launched by international organizations (IAEA-Vienna, NEA-OECD, NNDC-USA) such a project would have a powerful international support because of increasing world wide demand of atomic and nuclear data. Nuclear data are either structure and decay nuclear data or reaction nuclear data. The first class refers to nuclear state properties (masses, excitation energies, quantum numbers, lifetimes, etc.) as well as to their decay modes. Data from the second class refer to differential or integral cross sections. The paper presents comparatively the data obtainable at accelerators and reactors for the two above mentioned classes of nuclear data, particularly, the data required for building ADS (Accelerator Driven Systems

  13. Radiation exposure doses of employees in reactor facilities for test and research and under research and development stages, and in facilities for nuclear fuel refining, fabrication, reprocessing and usage

    International Nuclear Information System (INIS)

    (1) Radiation exposure doses in reactor facilities. The owners of reactor facilities are obliged by law to control the radiation exposure doses of the employees below the permissible levels. The data based on the reports made in this connection are given in tables for the fiscal year 1978 (from April 1978 to March 1979). It was revealed that the radiation exposure doses of the employees were far below the permissible levels. The distributions of exposure doses in Japan Atomic Energy Research Institute, Power Reactor and Nuclear Fuel Development Corporation and so on are presented for the whole year and the respective quarters. (2) Radiation exposure doses in facilities for nuclear fuel. The owners are similarly obliged to control radiation exposure. The data in this connection are given, and the doses were far below the permissible levels. The distributions in the private enterprises and so on are presented for the whole year. (J.P.N.)

  14. Feasibility study of the underwater neutron radiography facility using the University of Utah 100 kW TRIGA (UUTR) reactor

    International Nuclear Information System (INIS)

    The University of Utah 100 kW TRIGA (UUTR) reactor provides usable neutron yields for neutron radiography. Currently, UUTR reactor has three irradiators (Central, Pneumatic, and Thermal irradiators) and one Fast neutron Irradiation Facility (FNIF). These irradiators are very small so they are not suitable for neutron radiography. UUTR has three beam ports but they are not available due to the structure of the core. All sides of the core are occupied by FNIF, Thermal Irradiator, and three ion chambers. The only available position for underwater vertical beam port is on the top of the FNIF. There are two factors necessary to fulfill to be able to realize vertical underwater beam port: noninterruption to other facilities and radiation shielding. Designing the vertical beam port as movable ensures good access to the core and pool, while still providing a good neutron radiography environment. Keeping the top of the beam port below the surface of the pool the water represents biological shield. Neutron radiographs, with a simple setup of efficient neutron converters and digital camera systems, can produce acceptable resolution with an exposure time as short as a few minutes. It is important to validate the design with calculations before constructing the beam port. The design of the beam port is modeled using the MCNP5 transport code. A minimum of 105 neutrons/cm2-sec thermal neutron flux is required for high resolution neutron radiography. Currently, the UUTRIGA is in the process of upgrading its power from 100 kW to 250 kW. Upon the completion of the upgrading, the maximum neutron flux in the core will be ∼7x1012 neutrons/cm2-sec. This paper discusses a modeling and evaluation of capability for a neutron radiography facility. (author)

  15. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  16. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  17. Dosimetric evaluation of semiconductor detectors for application in neutron dosimetry and microdosimetry in nuclear reactor and radiosurgical facilities

    International Nuclear Information System (INIS)

    The main objective of this research is the dosimetric evaluation of semiconductor components (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB - Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation methodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established. (author)

  18. Parametric analysis of the neutron transmutation doping (NTD) facility at the Egyptian Second Research Reactor (ETRR-2)

    International Nuclear Information System (INIS)

    One of the most important facilities of the Egyptian Second Research Reactor (ETRR-2) is the silicon Neutron Transmutation Doping (NTD) Facility. This work is done to optimize the performance of the ETRR-2 NTD facility based on the standards and to develop a calculation line that can be used during its commercial operating phase. Calculation line methodology was developed using WIMS-D4 and CITVAP codes as well as nuclear data file libraries to perform core calculations needed to predict neutronic parameters at the two NTD locations. Axial flux profile was calculated and results were in good agreement with the experimental flux mapping. Based on the calculated variation of the thermal neutron flux axial distribution, it was proposed to displace the silicon ingot axially from its initially proposed position to minimize flux variation. Impact of control rod configuration was analyzed to maximize the neutron flux at the NTD two locations and the best control rod configuration is recommended. Thermal to fast neutron flux was also calculated and concluded that radiation damage caused by fast neutrons can safely be neglected. Flux depression across the ingot was estimated using the ANISN code. Finally, the SIMULINK module of the MATLAB code was used to evaluate the ingot activity due to different radioactive nuclides. The activity was found to reach its exempted limit after 2.79 days after its discharge from the irradiation channel. It was also concluded that the worse ETRR-2 scram will not affect the obtained Si doping and its resistivity

  19. The different facilities of the reactor PHENIX for radio isotope production and fission product burner

    International Nuclear Information System (INIS)

    During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (author)

  20. Teaching and Training at RA-6 Reactor and their Contribution to the Research Facilities Development

    International Nuclear Information System (INIS)

    The RA-6 reactor at the Bariloche Atomic Centre was designed mainly as a teaching tool. During its almost 21 years, after the first criticality, it was used as a support for several graduate and post graduate careers at the Balseiro Institute, depending on the Argentine National Atomic Energy Commission and the Cuyo National University. Besides these tasks, a big work was done in research and development using the synergy produced by the close relationship between the students and researchers. Main characteristics of the reactor are described. An outline of use for teaching and training is given. Research activities resulting from appropriate personnel management and students' cooperation that allowed achieving different grades of development are described. (author)

  1. Fresh and spent nuclear fuel repatriation from the IRT-2000 research reactor facility, Sofia, Bulgaria

    International Nuclear Information System (INIS)

    The IRT-2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned. (author)

  2. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    International Nuclear Information System (INIS)

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned

  3. Facility modification for severe accident mitigation at the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    As a part of the recently completed Safety Analysis Report (SAR) for the High Flux Isotope Reactor (HFIR), a detailed MELCOR model was developed for use in the analysis of selected events that lead to core damage. This detailed MELCOR model included the HFIR's reactor coolant system, dynamic confinement, building and yard piping, and a model of the outside waste collection storage tanks. Analyses of several accident sequences involving large breaks in the cold leg piping in the heat exchanger cells and pipe tunnel were conducted with the model. From the results of these simulations, a need for modification of the HFIR confinement was identified. These confinement modifications resulted in over a 12 hour delay in the release of primary system water (an any fission products it contains) outside of the HFIR confinement. This delay would provide valuable time to prevent or mitigate any releases to the environment due to a break in the primary system of the HFIR

  4. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  5. Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

    International Nuclear Information System (INIS)

    A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

  6. Operational experience on sodium deposits in KNK reactor and RSB test facility

    International Nuclear Information System (INIS)

    A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

  7. The role of neutron activation analysis technique Ex Industrial applications using the egyptian research reactor facilities

    International Nuclear Information System (INIS)

    This report covers several papers which deal with the industrial applications of the Neutron Activation Analysis Technique (NAAT) in Egypt. The applications include: exploration, mining, industrial environment and multielemental analysis of different materials, just for quality control, optimization, safety uses and help in improving the efficiency and economic evaluation. The technique principles, instrumentation, neutron irradiation facilities and experience of analysis are reviewed. Also, the current research activities using the ET-RR-1 facilities as well as a proposal for cold neutron applications in this field on the ET-RR-2 are given

  8. On results of the decommissioning of Georgian nuclear research reactor IRT by grouting and its conversion into a new low power nuclear facility

    International Nuclear Information System (INIS)

    The research nuclear reactor IRT- M belonged to the Institute of Physics of Georgian Academy of Sciences. Its operation came to an end in 1988. Taking into account the lack of financial and technical resources, the Institute of Physics issued decision to shut down reactor by grouting the radioactive low part of the reactor tank, where the reactor core is placed. It gave us the possibility to carry out decommissioning of the reactor using modest finances. Installing a new low-power facility (about 50 kW, nuclear fuel -20% enrichment by 235U) onto the free part of the reactor tank was also considered. The proposed facility is designed to carry out neutron-activation for producing of radionuclides for medicine, curing tumors by Boron Neutron Capture Therapy (BCNT) and others aims. Solving of all these tasks is very important for Georgia. The plan for dismantling of remaining cooling and other auxiliary systems of the reactor IRT-M has already been elaborated. All activities are performed within the frame of new IAEA TC project GEO/3/002. (author)

  9. Measurements of reactor-relevant electromagnetic effects with the FELIX facility

    International Nuclear Information System (INIS)

    Recent experiments with the FELIX (Fusion Electromagnetic Induction eXperiment) facility at Argonne National Laboratory (ANL) suggest that the expected electromagnetic forces and torques in a tokamak first wall, blanket, and shield (FWBS) system can be modelled by a single eddy current mode, with a simple characterization

  10. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  11. Application of 'FLUENT' to describe thermo-hydraulic processes in experimental facilities which model severe accident development in nuclear reactor

    International Nuclear Information System (INIS)

    The work is aimed at reviewing the applicability of the 'FLUENT v6.2' up-to-date software system for modeling such thermo hydraulic processes as boiling/condensation and melting/solidification concurrently taking place in multicomponent systems. The work presents an example of calculation modeling of processes taking place in experiments performed by National Nuclear Center of the Republic of Kazakhstan for research into final stages of an accident in nuclear reactor. Results of experimental works are used for nuclear facilities safety justification. Application of 'FLUENT v6.2' software system enables to reproduce sequence of events taking place in experiments and to forecast their development that is necessary for design of experiments and for results analysis as well. (author)

  12. Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

  13. Optimization of a neutron detector system for the fourier RTOF-diffractometer facility at the ETRR-1 reactor

    International Nuclear Information System (INIS)

    This work deals with optimization of a detector system for use with the fourier reverse time of flight (RTOF) diffractometer facility recently installed at the ETRR-1 reactor. The detector system of the diffractometer was designed for detecting neutrons scattered from the sample. It consists of 4 independent scintillation detector elements arranged according to the time focusing geometry in order to increase the luminosity of the diffractometer for the given resolution. Each of the detector elements is made of 1 mm thick 6 Li-glass scintillator (NE-912) whose surface area is 200 mm2. The present detector system ensures reliable identification and separation of thermal neutron pulses from fast neutrons and gamma- quanta. It was found from measurements with Fe sample that it is possible to make neutron diffraction measurement, within one hour using the present detector system and with resultion 0.5%. 6 FIGS

  14. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D2O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  15. Sorption purification of Pb-containing wash and spent decontamination solutions of reactor facilities with HLMC from radionuclides

    International Nuclear Information System (INIS)

    Consideration is given to sorption purification techniques of liquid radioactive wastes of spent wash and decontamination solutions (SWDS) of reactor facilities with heavy liquid metal coolants. Possibility of sorption purification of SWDS from radionuclides has been examined experimentally in respect to sorption of main radionuclides 137Cs and90Sr from model solutions. It is shown that none of natural adsorbents studied (tripoli, clinoptilolite, bentonite) provides necessary purification of model SWDS from radionuclides. For this reason, for developing the SWDS treatment technique and optimization of parameters of sorption-membrane recovery of radionuclides the process is recommended which consists of precipitation-filtration operations of lead low-soluble salts recovery or operation of dilution in combination with membrane-sorption operations of contaminating radionuclides removing

  16. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  17. Simple computational modeling for human extracorporeal irradiation using the BNCT facility of the RA-3 Reactor

    International Nuclear Information System (INIS)

    We present a simple computational model of the reactor RA-3 developed using Monte Carlo transport code MCNP. The model parameters are adjusted in order to reproduce experimental measured points in air and the source validation is performed in an acrylic phantom. Performance analysis is carried out using computational models of animal extracorporeal irradiation in liver and lung. Analysis is also performed inside a neutron shielded receptacle use for the irradiation of rats with a model of hepatic metastases.The computational model reproduces the experimental behavior in all the analyzed cases with a maximum difference of 10 percent. (author)

  18. Probabilistic evaluation of seismic isolation effect with respect to siting of a fusion reactor facility

    International Nuclear Information System (INIS)

    Annual failure probabilities of buildings and equipment were roughly evaluated for two fusion-reactor-like buildings, with and without seismic base isolation, in order to examine the effectiveness of the base isolation system regarding siting issues. The probabilities are calculated considering nonlinearity and rupture of isolators. While the probability of building failure for both buildings on the same site was almost equal, the function failures for equipment showed that the base-isolated building had higher reliability than the non-isolated building. Even if the base-isolated building alone is located on a higher seismic hazard area, it could compete favorably with the ordinary one in reliability of equipment

  19. Probabilistic evaluation of seismic isolation effect with respect to siting of a fusion reactor facility

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Masatoshi; Komura, Toshiyuki; Hirotani, Tsutomu [Shimizu Corp., Tokyo (Japan). Power and Energy Project Division; Ohkawa, Yoshinao; Akutsu, Youich [Japan Atomic Energy Research Inst., Ibaraki (Japan)

    1995-12-01

    Annual failure probabilities of buildings and equipment were roughly evaluated for two fusion-reactor-like buildings, with and without seismic base isolation, in order to examine the effectiveness of the base isolation system regarding siting issues. The probabilities are calculated considering nonlinearity and rupture of isolators. While the probability of building failure for both buildings on the same site was almost equal, the function failures for equipment showed that the base-isolated building had higher reliability than the non-isolated building. Even if the base-isolated building alone is located on a higher seismic hazard area, it could compete favorably with the ordinary one in reliability of equipment.

  20. EURAC: the JRC proposal for an European fusion reactor materials test and development facility

    International Nuclear Information System (INIS)

    For the last 7 years we examined the use of a Spallation Neutron Source (SNS) as an altenative European Option to FMIT. For an optimized spallation neutron source design we find now for the same beam power the following design parameters: - Linear Accelerator: 600 MeV, 6 m-A-proton beam on liquid lead target - irradiation parameters: 320 dpa/year in 20 cm3 or 274 dpa/year in 31.5 cm3 6 -1 sec-1 in order to simulate the Pulsed Mode of Tokamak Power Reactors. The deflected beam can be used for other experiments

  1. Analysis of a laser-initiated, inertially-confined reactor for a fusion engineering research facility (LA FERF)

    International Nuclear Information System (INIS)

    With the increasing optimism for the viability of laser-induced, inertially-confined fusion-reactor power plants, comes an increased interest in a high flux 14-MeV neutron generator. This generator can be used to investigate 14-MeV neutron damage in first-wall material, and various neutron energy direct conversion schemes. The associated charged particles can be used to investigate various first-wall designs proposed for laser-fusion reactor systems as well as the direct conversion of charged particle energy by expansion of a fusion fireball against a magnetic field imposed from outside the neutron generator first wall. A systematic parameter study is made on a laser-initiated, inertially-confined reactor for a fusion engineering research facility (LA FERF). The parameters investigated are the variations of plant cost, plant power, and plant performance as a function of the fusion gain and power input to the laser system. Design point envelopes are presented for the laser input power and fusion gain (defined as the ratio of the thermonuclear power to power into the laser system) based on reference plant costs, powers and performance. A design point was selected (based on technology available in the not-too-distant future) which produces a flux of 1.8 x 1018 n/m2 . s over an experimental area of approximately 1 m2 for a cost of 160 megadollars using a power of 140 MVA. A comparison is made with a current mirror FERF study proposed for use in CTR engineering studies. The LA FERF could be built in the mid 1980's using near-term short pulse CO2 laser technology, producing thermonuclear power of 6 to 20 MW with fusion gains of approximately 0.1

  2. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  3. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    Energy Technology Data Exchange (ETDEWEB)

    Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

    1999-08-01

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  4. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  5. Measurement of neutron spectrum at irradiation facilities of Kyoto University Reactor, KUR

    International Nuclear Information System (INIS)

    By unfolding multi-foil activation data with NEUPAC code, neutron energy spectra at the irradiation facilities (hydraulic conveyoy tube, long irradiation plug, graphite thermal neutron column, low temperature loop and pneumatic tubes) of KUR have been obtained in the energy region from 0.01 eV to 16.4 MeV. Transport calculations by ANISN code have been also made, and agreed with the measured spectra, in general. By using the neutron spectra measured above, the spectrum-averaged cross section for the 54Fe(n,p)54Mn reaction has been obtained as a basic datum to estimate the neutron fluence at the irradiation facilities. Neutron spectra at the low temperature loop and the pneumatic tubes have been also measured for the special core of KUR, whose fuel configuration was changed so that the exposure tube E-4 (low temperature loop) could directly see the fuel elements. (author)

  6. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  7. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    International Nuclear Information System (INIS)

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  8. FFTF (FAST FLUX TEST FACILITY) REACTOR CHARACTERIZATION PROGRAM ABSOLUTE FISSION RATE MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    FULLER JL; GILLIAM DM; GRUNDL JA; RAWLINS JA; DAUGHTRY JW

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  9. FFTF (Fast Flux Test Facility) Reactor Characterization Program: Absolute Fission-rate Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, J.L.; Gilliam, D.M.; Grundl, J.A.; Rawlins, J.A.; Daughtry, J.W.

    1981-05-01

    Absolute fission rate measurements using modified National Bureau of Standards fission chambers were performed in the Fast Flux Test Facility at two core locations for isotopic deposits of {sup 232}Th, {sup 233}U, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Pu. Monitor chamber results at a third location were analyzed to support other experiments involving passive dosimeter fission rate determinations.

  10. Scabol and LOFW Experiments in A Scaled-Down IET Facility of REX-10 Reactor

    International Nuclear Information System (INIS)

    This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10

  11. Surplus Facilities Management Program. Post remedial action survey report for the Sodium Reactor Experiment (SRE) facility, Santa Susana Field Laboratories, Rockwell International, Ventura County, California

    International Nuclear Information System (INIS)

    Decontamination of the Sodium Reactor Experiment (SRE) began in 1976 and was completed in 1982. In view of the concurrent and post-remedial-action surveys, the following conclusions can be stated. All the buildings and areas included in this decommissioning project have been decontaminated to below the limits specified in the draft ANSI Standard N13.12 and the NRC Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for By-Product, Source, or Special Nuclear Material, dated July 1982. Radioactive contamination was found in appropriate access points of the sanitary sewer and storm drain systems included within the boundaries of this decommissioning project. One sample indicated a 90Sr concentration dissolved in the water of approximately half the recommended water concentration for controlled areas and approximately 15 times the recommended water concentration for uncontrolled areas as stated in DOE-5480.1 Chg. 6, Chapter XI. Therefore, the interior inaccessible surfaces of these systems must be considered contaminated in accordance with statements found in the NRC Regulatory Guidelines issued in July 1982. Effluent from the outfall of this drain system must also be considered as being potentially contaminated. 1 reference, 32 figures, 8 tables

  12. Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  13. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  14. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    Energy Technology Data Exchange (ETDEWEB)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-04-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ``flooded cavity``, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications.

  15. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    International Nuclear Information System (INIS)

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  16. Development of a methodology for safety classification on a non-reactor nuclear facility illustrated using an specific example

    International Nuclear Information System (INIS)

    To realize the safety of personnel and environment systems and components of nuclear facilities are classified according to their potential danger into safety classes. Based on this classification different demands on the manufacturing quality result. The objective of this work is to present the standardized method developed by NUKEM Technologies Engineering Services for the categorization into the safety classes restricted to Non-reactor nuclear facilities (NRNF). Exemplary the methodology is used on the complex Russian normative system (four safety classes). For NRNF only the lower two safety classes are relevant. The classification into the lowest safety class 4 is accordingly if the maximum resulting dose following from clean-up actions in case of incidents/accidents remains below 20 mSv and the volume activity restrictions of set in NRB-99/2009 are met. The methodology is illustrated using an example. In short the methodology consists of: - Determination of the working time to remove consequences of incidents, - Calculation of the dose resulting from direct radiation and due to inhalation during these works. The application of this methodology avoids over-conservative approaches. As a result some previously higher classified equipment can be classified into the lower safety class.

  17. General arrangement design optimization of emergency response facility (TSC, OSC) in Korea next generation reactor (APR-1400)

    International Nuclear Information System (INIS)

    The accident at the Three Mile Island (TMI) led to install some data processing and display equipment to assist control room personnel in rapidly evaluating the safety status. And also to place in the plant for providing operators with technical support and an emergency response center for radiological environmental assessments and determination of recommended public protective action during emergency. In practice, most of the countries possessing nuclear power plants including USA have partially or wholly adopted US NRC regulations and guidelines for Emergency Response Facility(ERF). Also the Korea nuclear power plants are implementing or operating ERF and SPDS after analyzing US NRC regulations and guideline since TMI accident. So this paper first been reviewed Korea Regulations, US NRC published codes and standards related to ERF (TSC/OSC). Finally this paper is described the design optimization of general arrangement in emergency response facility to improve emergency response capability in Korea Next Generation Reactor(APR-1400), which are best suitable for our domestic situation and also enhance the emergency response capability of ERF

  18. Establishment and prioritization of relevant factors to the safety of fuel cycle facilities non reactor through dynamics archetypes evaluation

    International Nuclear Information System (INIS)

    The present work aims to establish and prioritize factors that are important to the safety of nuclear fuel cycle facilities in order to model, analyze and design safety as a physical system, employing systemic models in an innovative way. This work takes into consideration the fact that models that use adaptations of methodologies for nuclear reactors will not properly work due to the specificities of fuel cycle facilities. Based on the fundamentals of the theory of systems, the four levels of system thinking, and the relationship of eight socio technical factors, a mental model has been developed for safety management in the nuclear fuel cycle context. From this conceptual model, safety archetypes were constructed in order to identify and highlight the processes of change and decision making that allow the system to migrate to a state of loss of safety. After that, stock and flow diagrams were created so that their behavior could be assessed by the system's dynamics. The results from the analysis using the model that simulates the dynamic behavior of the variables (socio technical factors) indicated, as expected, that the system's dynamics proved to be an appropriate and efficient tool for modeling fuel cycle safety as an emergent property. (author)

  19. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  20. A reverse time of flight analyzer facility at the ETRR-1 reactor

    International Nuclear Information System (INIS)

    The present work deals both with the theory and performance of a reverse-time-of-flight (RTOF) analyzer designed to analyze pulses emitted from a fourier chopper recently put into operation at the ETRR-1 reactor. The RTOF analyze was found to be adequate for use with pick up pulses from the fourier chopper which operates following a frequency window suitable for rotation rates from 0-9000 rpm; synchronically with neutron pulses from a 6 Li glass detector set at time focusing geometry for scattering angle 20=90 degree. It was possible, with the present RTOF analyzer to obtain diffraction patterns at neutron wavelength range between 1 - 4 A within a resolution = 0.5%. 8 FIGS