Reactor Sharing at Rensselaer Critical Facility

International Nuclear Information System (INIS)

D. Steiner, D. Harris, T. Trumbull

2006-01-01

This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

Advanced reactor experimental facilities

International Nuclear Information System (INIS)

Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

2010-01-01

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

Annual report 1986 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1987-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of Current research during 198? of the departments radiochemistry, radiation chemistry, radiation physics, and reactor physics. (H.W.). refs.; figs.; tabs

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1980-09-01

Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

International Nuclear Information System (INIS)

Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

2014-03-01

The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

Annual report 1984 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1985-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1984 of the departments radiochemistry, radiation chemistry, radiation physics, reactor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

Annual report on JEN-1 reactor

International Nuclear Information System (INIS)

Montes, J.

1972-01-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

Facility Effluent Monitoring Plan for the N Reactor

International Nuclear Information System (INIS)

Watson, D.J.; Brendel, D.F.; Shields, K.D.

1991-11-01

A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The primary purpose of the N Reactor Facility Effluent Monitoring Plan (FEMP), during standby, is to ensure that the radioactive effluents are properly monitored and evaluated for compliance with the applicable DOE orders and regulatory agencies at the federal, state, and local levels. A secondary purpose of the FEMP is to ensure that hazardous wastes are not released, in liquid effluents, to the environment even though the potential to do so is extremely low. The FEMP is to provide a monitoring system that collects representative samples in accordance with industry standards, performs analyses within stringent quality control (QC) requirements, and evaluates the data through the use of comparative analysis with the standards and acceptable environmental models

Reactor incident status 1981 annual report

International Nuclear Information System (INIS)

Kiser, S.H.

1982-01-01

Reactor Incident followup action is summarized through periodic status reports. This annual report summarizes action taken or anticipated for Reactor Incidents through December 1981. Incidents for which action has been completed, have been deleted from the report. Quarterly addende will update the report by tabulating incidents for each three month period through the coming year. The report consists of a part for the P, K, and C Reactors. Each reactor part is divided into three sections: Further Technical Analysis or Followup Needed; Funding and/or Implementation Needed; and No Further Technical Analysis Anticipated

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1993-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor engineering department annual report

International Nuclear Information System (INIS)

1990-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Emergency reactor core cooling facility

International Nuclear Information System (INIS)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

1996-01-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

Emergency reactor core cooling facility

Energy Technology Data Exchange (ETDEWEB)

Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

1996-11-01

The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

RA Research reactor, Annual report 1988

International Nuclear Information System (INIS)

Sotic, O.

1988-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

International Nuclear Information System (INIS)

1989-01-01

The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

International Nuclear Information System (INIS)

1994-01-01

The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

Estimated Annual Maintenance Costs for Educational Facilities in Eritrea

DEFF Research Database (Denmark)

Vagnby, Bo Hellisen

Global estimated annual costs for undertaking basic maintenance of all educational facilities in Eritrea.......Global estimated annual costs for undertaking basic maintenance of all educational facilities in Eritrea....

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

International Nuclear Information System (INIS)

Raddatz, C.T.

1992-04-01

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)

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

International Nuclear Information System (INIS)

Raddatz, C.T.

1991-07-01

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

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

International Nuclear Information System (INIS)

Raddatz, C.T.

1993-07-01

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)

Facilities of fuel transfer for nuclear reactors

International Nuclear Information System (INIS)

Wade, E.E.

1977-01-01

This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

RA Research nuclear reactor - Annual report 1987

International Nuclear Information System (INIS)

1987-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Research nuclear reactor RA - Annual Report 1989

International Nuclear Information System (INIS)

Sotic, O.

1989-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

Reactor facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

1997-01-01

The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

International Nuclear Information System (INIS)

Salazar, M.; Elder, J.

1992-08-01

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

Nursing Facility Initiative Annual Report

Data.gov (United States)

U.S. Department of Health & Human Services — This annual report summarizes impacts from the Initiative to Reduce Avoidable Hospitalizations among Nursing Facility Residents in 2014. This initiative is designed...

Present status of decommissioning in the Musashi Reactor Facility (4)

International Nuclear Information System (INIS)

Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

2012-01-01

The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

RB research nuclear reactor, Annual report for 1981

International Nuclear Information System (INIS)

Markovic, H.; Sotic, O.; Pesic, M.; Vranic, S.; Zivkovic, B.; Bogdanovic, M.; Petronijevic, M.

1981-01-01

The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

International Nuclear Information System (INIS)

1995-01-01

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

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

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

Annual report 1987-1988

International Nuclear Information System (INIS)

1988-01-01

The paper presents: the annual report 1987/8 of the University Research Reactor, Risley, United Kingdom. The report contains a description of the current research programme using reactor produced isotopes and other reactor facilities. A summary of the work carried out by the Activation Analysis Service during 1987/8 is also given, along with the Reactor Teaching Programme, and the reactor operation and site safety arrangements. (U.K.)

Annual report 1983 Interfacultary Reactor Institute

International Nuclear Information System (INIS)

1984-01-01

In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1983 of the departments radiochemistry, radiation chemistry, radiation physics, reacotor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

Nuclear reactor containing facility

International Nuclear Information System (INIS)

Hidaka, Masataka; Murase, Michio.

1994-01-01

In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

Reactor cold neutron source facility, the first in Japan

International Nuclear Information System (INIS)

Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

1986-01-01

In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

Design requirements for new nuclear reactor facilities in Canada

International Nuclear Information System (INIS)

Shim, S.; Ohn, M.; Harwood, C.

2012-01-01

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)

Neutron beam facilities at the replacement research reactor

International Nuclear Information System (INIS)

Kennedy, S.

1999-01-01

Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

Irradiation Facilities at the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2005-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

Annual report 1986 of Society for Reactor Safety (GRS)

International Nuclear Information System (INIS)

1987-01-01

The overall development of Society for Reactor Safety (GRS) is indicated by its annual volume of achievement. After initially high growth rates that achievement has come to stand at some 62 million marks or so over the last years. GRS today employs some 450 staff, of which 325 are scientific-technical experts representing altogether nearly 4000 man-years of experience in nuclear technology. The report presents essential results from the main points of activity of GRS over 10 years. In accordance with the aim of the institution they primarily cover projects in the sectors safety and radiation protection for nuclear facilities. Since 1985, since the extension of its contract of associates, GRS has also increasingly been concerned with safety questions in non-nuclear technology, concentrating mainly on the transfer of existing methods. (orig./DG) [de

Risk management activities at the DOE Class A reactor facilities

International Nuclear Information System (INIS)

Sharp, D.A.; Hill, D.J.; Linn, M.A.; Atkinson, S.A.; Hu, J.P.

1993-01-01

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

Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

International Nuclear Information System (INIS)

Kawahara, Toshio; Matsushita, Tadashi

1977-01-01

The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

Progress report concerning safety research for nuclear reactor facilities

International Nuclear Information System (INIS)

1978-01-01

Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

Research nuclear reactor RA - Annual report 1992

International Nuclear Information System (INIS)

Sotic, O.

1992-12-01

Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

RB research nuclear reactor, Annual report for 1984, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

1984-01-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

International Nuclear Information System (INIS)

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1993-March 31, 1994

International Nuclear Information System (INIS)

1994-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-10-01

This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

International Nuclear Information System (INIS)

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Nuclear Safety Research and Facilities Department. Annual report 1999

International Nuclear Information System (INIS)

Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

2000-04-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

Nuclear Safety Research and Facilities Department annual report 1997

International Nuclear Information System (INIS)

Majborn, B.; Aarkrog, A.; Brodersen, K.

1998-04-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department's research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

Nuclear Safety Research and Facilities Department annual report 1998

International Nuclear Information System (INIS)

Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

1999-04-01

The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

International Nuclear Information System (INIS)

1988-01-01

A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

Twenty-ninth annual progress report of the Pennsylvania State University Breazeale Nuclear Reactor, July 1, 1983-June 30, 1984

International Nuclear Information System (INIS)

Levine, S.H.; Totenbier, R.E.

1984-07-01

The twenty-ninth annual progress report of the operation of the Pennsylvania State University Breazeale Reactor is submitted in accordance with the requirements of Contract DE-AC02-76ER03409 with the United States Department of Energy. This report also provides the University administration with a summary of the operation of the facility for the past year

Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

International Nuclear Information System (INIS)

1995-01-01

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

Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

International Nuclear Information System (INIS)

Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

1996-01-01

Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

Nuclear Safety Research and Facilities Department. Annual report 1999

Energy Technology Data Exchange (ETDEWEB)

Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E. [eds.

2000-04-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

Nuclear Safety Research and Facilities Department annual report 1999

DEFF Research Database (Denmark)

Majborn, B.; Damkjær, A.; Jensen, Per Hedemann

2000-01-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department´s research and development activities were organized in two research programmes: "Radiation Protection and Reactor Safety" and"Radioecology and Tracer Studies". The nuclear...... facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are includedtogether with a summary of the staff´s participation in national and international committees....

Nuclear Safety Research and Facilities Department annual report 1997

Energy Technology Data Exchange (ETDEWEB)

Majborn, B.; Aarkrog, A.; Brodersen, K. [and others

1998-04-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department`s research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 11 tabs., 39 ills.; 74 refs.

Nuclear Safety Research and Facilities Department annual report 1998

Energy Technology Data Exchange (ETDEWEB)

Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E

1999-04-01

The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department`s research and development activities were organized in two research programmes: `Radiation Protection and Reactor Safety` and `Radioecology and Tracer Studies`. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au)

Nuclear Safety Research and Facilities department annual report 1996

International Nuclear Information System (INIS)

Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Heydorn, K.; Oelgaard, P.L.

1997-04-01

The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1996. The Department's research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au) 2 tabs., 28 ills

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

International Nuclear Information System (INIS)

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1994 - March 31, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

Energy Technology Data Exchange (ETDEWEB)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

International Nuclear Information System (INIS)

Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

1998-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

Reactor engineering department annual report. April 1, 1995 - March 31, 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

RB research nuclear reactor, Annual report for 1983, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

1983-01-01

The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

Standard irradiation facilities for use in TRIGA reactors

International Nuclear Information System (INIS)

Kolbasov, B.N.; Luse, R.A.

1972-01-01

The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

International Working Group on Fast Reactors Second Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1969-01-01

The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

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

Annual report ''nuclear safety in France''

International Nuclear Information System (INIS)

2001-01-01

This document is the 2001 annual report of the French authority of nuclear safety (ASN). It summarizes the highlights of the year 2000 and details the following aspects: the nuclear safety in France, the organization of the control of nuclear safety, the regulation relative to basic nuclear facilities, the control of facilities, the information of the public, the international relations, the organisation of emergencies, the radiation protection, the transport of radioactive materials, the radioactive wastes, the PWR reactors, the experimental reactors and other laboratories and facilities, the nuclear fuel cycle facilities, and the shutdown and dismantling of nuclear facilities. (J.S.)

Advanced Test Reactor National Scientific User Facility Partnerships

International Nuclear Information System (INIS)

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

2012-01-01

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 Wisconsin

RB research nuclear reactor - Annual report for 1986, I - III

International Nuclear Information System (INIS)

Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

1987-01-01

This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

Operation and maintenance of the RB reactor, Annual report for 1977

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1977-01-01

The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators

The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

International Nuclear Information System (INIS)

Vail, J.

1992-01-01

Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

International Nuclear Information System (INIS)

Forsbacka, Matt

1990-01-01

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)

Feasibility study to develop BNCT facility at the Indonesian research reactor

International Nuclear Information System (INIS)

Hastowo, H.

2001-01-01

A survey on the Indonesian research reactors and its supporting facilities has been done in order to check the possibility to install BNCT facility. Oncologists from several hospitals have been informing about the BNCT treatment for tumours and they give a positive response to support utilisation of the BNCT facility. Several aspects required to support the BNCT treatment have also been identified and related activities on that matter soon will be initiated. The interim result in our survey indicated that utilisation of the 30 MW Multipurpose reactor would not be possible from the technical point of view. Further study will be concentrated on the TRIGA reactor and an epithermal neutron beam facility at the thermal column of this reactor will be designed for further work. (author)

Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

Energy Technology Data Exchange (ETDEWEB)

Gutmann, H

1972-06-05

The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

Operation and maintenance of the RB reactor, Annual report for 1978

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1978-01-01

The annual report for 1978 includes the following: utilization of the RB reactor; producing the new safety report; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Four annexes to this report are concerned with: operation of the reactor at higher power levels; performance of the instrumentation, radiation doses during operation; gamma radiation doses after reactor shutdown; properties of the neutron converter (optimization of the rector-converter coupling; maintenance of the reactor equipment; purchase of new equipment

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

1997-01-01

The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

Annual report of Nuclear Engineering Research Laboratory, University of Tokyo in fiscal 1993

International Nuclear Information System (INIS)

1994-08-01

In this annual report, the activities of research and education, the state of operation of research facilities and others in fiscal year 1993 are summarized. Four main research facilities are the fast neutron source reactor 'Yayoi', the electron linear accelerator, the basic experiment facility for nuclear fusion reactor blanket design and the heavy irradiation research facility. The reactor and the accelerator are for the joint utilization by all universities in Japan, the blanket is used by the Faculty of Engineering, and the HIT is for the joint utilization in University of Tokyo. In fiscal year 1993, the installation of the fast neutron science research facility was approved. In this annual report, the management and operation of the above research facilities are described, and the research activities, the theses for doctorate and graduation theses of teachers, are summarized. (K.I.)

Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

Energy Technology Data Exchange (ETDEWEB)

Montes, J

1972-07-01

In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Nuclear Reactor Laboratory annual report, fiscal year 1981-1982

International Nuclear Information System (INIS)

Cashwell, R.J.

1982-01-01

Information related to the use of the UWNR reactor is presented concerning instructional use by the Nuclear Engineering Department; reactor sharing program; utility personnel training; sample irradiations and neutron activation analysis services; changes in personnel, facility, and procedures; and results of surveillance tests

Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1979-09-01

Research activities in the Division of Reactor Engineering in fiscal 1978 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

Operation and maintenance of the RB reactor, Annual report for 1979

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.; Petronijevic, M.

1979-01-01

The annual report for 1979 includes the following: utilization of the RB reactor; accident analysis; dosimetry and radiation protection; description of the reactor equipment status; reactor staff; financial data. Six Annexes to this report are concerned with: scientific program orientation; producing the new safety report; maintenance of the reactor components and equipment; purchase of new equipment; financial report; and program for training the staff of the Krsko NPP

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.

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

International Nuclear Information System (INIS)

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

2008-01-01

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: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

Emergency reactor container cooling facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Matsumoto, Tomoyuki.

1992-01-01

The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

On present situation of radioactive waste management and exposure of workers in nuclear reactor facilities for commercial power generation in fiscal 1988

International Nuclear Information System (INIS)

1989-01-01

The article summarizes the contents of some reports including the Report on Radiation Management in 1988 that were submitted by the operators of nuclear reactor facilities for commercial power generation according to the requirements specified in the Law Concerning Regulation on Nuclear Material, Nuclear Fuel and Nuclear Reactor. According to these reports, the annual radiation release in all nuclear power generation plants was well below the radiation release limits set up in the report 'On Guidelines for Target Dose in Areas around Light Water Reactor Facilities for Power Generation'. Data submitted also show that there are no significant problems with the management of radioactive solid waste. In all nuclear generation plants, the personal exposure of workers is below the permissible exposure dose specified in law. The Agency of Natural Resources and Energy is planned to further promote the development of advanced techniques for automatization and remote control of light water reactors and to provide effective guidance to electrical contractors for positive radiation management. (N.K.)

The 'MELUSINE' reactor at Grenoble, France, and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the MELUSINE 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 six information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities

Emergency facility control device for nuclear reactor

International Nuclear Information System (INIS)

Ikehara, Morihiko.

1981-01-01

Purpose: To increase the reliability of a nuclear reactor by allowing an emergency facility to be manually started and stopped to make its operation more convenient and eliminate the possibility of erroneous operation in an emergency. Constitution: There are provided a first water level detector for detecting a level lower than the first low water level in a reactor container and a second water level detector for detecting a level lower than the second low water level lower than the first low water level, and an emergency facility can be started and stopped manually only when the level is higher than the second low water level, but the facility will be started regardless of the state of the manual operation when the level is lower than the second low water level. Thus, the emergency facility can be started by manual operation, but will be automatically started so as to secure the necessary minimum operation if the level becomes lower than the second low water level and the stopping operation thereafter is forgotten. (Kamimura, M.)

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Kourachenkov, A.V.

1998-01-01

The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

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

International Nuclear Information System (INIS)

Alger, D.M.

1987-01-01

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 6-7x10 14 n/cm 2 -s in the central flux trap and beamport source fluxes of up to 1.2x10 14 n/cm 2 -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)

Annual report of Nuclear Engineering Research Laboratory, University of Tokyo in fiscal 1991

International Nuclear Information System (INIS)

1992-07-01

In this annual report, the activities of education and research, the state of operation of research facilities and others in Nuclear Engineering Research Laboratory, University of Tokyo in fiscal year 1991 are summarized. In this Laboratory, there are four large research facilities, that is, the fast neutron source reactor 'Yayoi', the electron beam linac, the nuclear fusion reactor blanket experiment facility and the heavy irradiation research facility. Those are used for carrying out education and research in the wide fields of nuclear engineering, and are offered also for joint utilization. The results of the research by using respective research facilities were summarized in separate reports. In this annual report, the course of the management and operation of respective research facilities is described, and the research activities, the theses for doctorate and graduation theses of the teachers, personnel and graduate students in the Laboratory are summarized. In the research, those on first wall engineering for fusion reactors, fuel cycle engineering, electromagnetic structure engineering, AI and robotics, quantum beam engineering, new type reactor design and so on are included. (K.I.)

Study of fast reactor safety test facilities. Preliminary report

International Nuclear Information System (INIS)

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

The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

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

The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

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

Nuclear reactor facility

International Nuclear Information System (INIS)

Wampole, N.C.

1978-01-01

In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

Environmental assessment for the deactivation of the N Reactor facilities. Revision 1

International Nuclear Information System (INIS)

1994-11-01

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. The N Reactor facilities are currently in a surveillance and maintenance program, and will eventually be decontaminated and decommissioned (D and D). Operation and maintenance of the facilities resulted in conditions that could adversely impact human health or the environment if left as is until final D and D. The Proposed Action would deactivate the facilities to remove the conditions that present a potential threat to human health and the environment and to reduce surveillance and maintenance requirements. The action would include surveillance and maintenance after deactivation. Deactivation would take about three years and would involve about 80 facilities. Surveillance and maintenance would continue until final D and D, which is expected to be complete for all facilities except the N Reactor itself by the year 2018

Department of Nuclear Safety Research and Nuclear Facilities annual report 1995

International Nuclear Information System (INIS)

Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Jacobsen, U.; Oelgaard, P.L.

1996-03-01

The report presents a summary of the work of the Department of Nuclear Safety Research and Nuclear Facilities in 1995. The department's research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au) 5 tabs., 21 ills

Department of Nuclear Safety Research and Nuclear Facilities annual report 1995

Energy Technology Data Exchange (ETDEWEB)

Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Jacobsen, U.; Oelgaard, P.L. [eds.

1996-03-01

The report presents a summary of the work of the Department of Nuclear Safety Research and Nuclear Facilities in 1995. The department`s research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 5 tabs., 21 ills.

The neutron beam facility at the Australian replacement research reactor

International Nuclear Information System (INIS)

Hunter, B.; Kennedy, S.

1999-01-01

Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

1997-01-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

Coupling of AST-500 heating reactors with desalination facilities

Energy Technology Data Exchange (ETDEWEB)

Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

1997-09-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

Effluent releases at the TRIGA reactor facility

Energy Technology Data Exchange (ETDEWEB)

Whittemore, W L [General Atomic Co., San Diego, CA (United States)

1974-07-01

The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

20. Annual report. OECD Halden reactor project. 1979

International Nuclear Information System (INIS)

1981-01-01

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

Training and research reactor facility longevity extension program

International Nuclear Information System (INIS)

Carriveau, G.W.

1991-01-01

Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

Experimental facilities for Generation IV reactors research

International Nuclear Information System (INIS)

Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

2013-06-01

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

Intense neutron irradiation facility for fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

General considerations for neutron capture therapy at a reactor facility

International Nuclear Information System (INIS)

Binney, S.E.

2001-01-01

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)

RA Research reactor, Annual report 1971

International Nuclear Information System (INIS)

Milosevic, D. et al.

1971-12-01

During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

State of the art of nuclear facilities with organic cooled reactors

International Nuclear Information System (INIS)

Brede, O.

1984-01-01

USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

International Nuclear Information System (INIS)

Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

2012-01-01

A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

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

OSU Reactor Sharing Program FY 1995 annual report

International Nuclear Information System (INIS)

Higginbotham, J.F.

1996-10-01

This is the annual report of the activities supported under the Oregon State University Reactor Sharing Program, award number DE-FG06-NE38137. The beginning date for the award was September, 30, 1995 and the end date was September 29, 1996. Work conducted under this award is internally administered at the Radiation Center through a project tasking system. This allows for excellent quality control for the work which is performed from the point of initial contact, through the reactor application, project report generation and financial accounting. For the current fiscal year, FY95, the total cost of the reactor sharing program, including Radiation Center contributions, was $66,323.20 of which $40,000.00 was supplied by the DOE Reactor Sharing Program. The details of individual project costs is given in Table 1. The work performed for the individual projects are described in the brief work descriptions given in Table 2

Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

International Nuclear Information System (INIS)

Harvey, R.S.

1983-01-01

An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

Advanced Test Reactor National Scientific User Facility

International Nuclear Information System (INIS)

Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

2011-01-01

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.

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.

University of Wisconsin, Nuclear Reactor Laboratory. Annual report, 1985-1986

International Nuclear Information System (INIS)

Cashwell, R.J.

1986-01-01

Operational activities for the reactor are described concerning nuclear engineering classes from the University of Wisconsin; reactor sharing program; utility personnel training; sample irradiations and neutron activation analysis; and changes in personnel, facility, and procedures. Results of surveillance tests are presented for operating statistics and fuel exposure; emergency shutdowns and inadvertent scrams; maintenance; radioactive waste disposal; radiation exposures; environmental surveys; and publications and presentations on work based on reactor use

Imperial College Reactor Centre annual report. 1984

International Nuclear Information System (INIS)

1985-01-01

The following new equipment is noted; for atomic absorption spectrometry to supplement the neutron activation analysis, and an additional nuclear data analysis system to improve the quality and speed of the service to users of the Centre's facilities. Users include undergraduates from the University of London, outside bodies such as the British Musueum, as well as departments of Colleges of the University of London. The reactor lost only three days through failures or faults. Two replacement fuel elements were put into the reactor during the year. The report contains brief accounts of 34 research programmes at the Centre. (U.K.)

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.

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

International Nuclear Information System (INIS)

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

Research nuclear reactor RA - Annual Report 1994

International Nuclear Information System (INIS)

Sotic, O.

1994-12-01

Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Activities related to improvement of Russian project were continued in 1994. Control and maintenance of the reactor components was done regularly and efficiently. Extensive repair of the secondary coolant loop is almost finished and will be completed in the first part of 1995 according to existing legal procedures and IAEA recommendations. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 47 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

Low temperature irradiation facility at Kyoto University Reactor (KUR)

International Nuclear Information System (INIS)

Atobe, Kozo; Okada, Moritami; Yoshida, Hiroyuki; Kodaka, Hisao; Miyata, Kiyomi.

1977-01-01

A new refrigeration system has been substituted to the low temperature irradiation facility at KUR instead of the previous one, since April in 1975. The model 1204 CTi He liquifier was designed to be modified for the refrigerator with the capacity of 30 watts at 10 K. The refrigeration capacity of 38 watts at 10 K was defined using a special cryostat and transfer-tubes, and the lowest temperature of about 18 K was measured using the irradiation loop without reactor operation. The reconstructed facility enables us to hold the many specimens simultaneously in the sample chamber of the irradiation loop at about 25 K during reactor operation of 5 MW. The irradiation dose has been reached about 6.6 x 10 16 n sub(f)/cm 2 and 6.1 x 10 17 n sub(th)/cm 2 with the normal reactor operation cycle of up to 77 hours. The stable operation condition of the machine and the special safety system for the refrigeration system enable us to maintain easily the facility with a constant operation condition for such a long time irradiation. Many kinds of low temperature neutron irradiation experiments are carried out using the facility, which techniques are partially reported. (auth.)

Decommissioning of reactor facilities (2). Required technology

International Nuclear Information System (INIS)

Yanagihara, Satoshi

2014-01-01

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

Brookhaven Reactor Experiment Control Facility, a distributed function computer network

International Nuclear Information System (INIS)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

Nuclear reactors built, being built, or planned: 1987

International Nuclear Information System (INIS)

1988-06-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1987. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually for Washington headquarters and field offices of DOE; from the US Nuclear regulatory Commission; from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The major change in this revision involves the data related to shutdown and dismantled facilities. Because this information serves substantially different purposes, it has been accumulated in a separate section, ''Reactors and Facilities Shutdown or Dismantled.'' Cancelled reactors or reactors whose progress has been terminated at some stage before operation are included in this section

ATR National Scientific User Facility 2015 Annual Report

Energy Technology Data Exchange (ETDEWEB)

Robertson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, J. Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ogden, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cole, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knight, Collin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teysseyre, Sebastien [Idaho National Lab. (INL), Idaho Falls, ID (United States); Benson, Jeff [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidrich, Brenden [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jackson, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bean, Lindy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

This is the 2015 Annual Report for the Nuclear Science User Facility. This report includes information on university-run research projects along with a description of the program and the capabilities offered researchers.

Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

International Nuclear Information System (INIS)

De, S.K.; Karhadkar, C.G.

2017-01-01

An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

Civaux nuclear facilities. 2009 annual report

International Nuclear Information System (INIS)

2010-01-01

This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Civaux, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

Chooz nuclear facilities. 2009 annual report

International Nuclear Information System (INIS)

2010-01-01

This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Chooz, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

Brennilis nuclear facilities. 2009 annual report

International Nuclear Information System (INIS)

2010-01-01

This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Brennilis, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

Annual report of department of research reactor, 1992

International Nuclear Information System (INIS)

1993-12-01

The department of research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1992 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as well as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

1985. Annual progress report

International Nuclear Information System (INIS)

1986-01-01

This annual progress report of the CEA Protection and Nuclear Safety Institut outlines a description of the progress made in each sections of the Institut Research activities of the different departments include: reactor safety analysis, fuel cycle facilities analysis; and associated safety research programs (criticality, sites, transport ...), radioecology and environmental radioprotection techniques; data acquisition on radioactive waste storage sites; radiation effects on man, studies on radioprotection techniques; nuclear material security including security of facilities, security of nuclear material transport, and monitoring of nuclear material management; nuclear facility decommissioning; and finally the public information [fr

Engineering evaluation/cost analysis for the 105-DR and 105-F Reactor facilities and ancillary facilities

International Nuclear Information System (INIS)

Coenenberg, E.T.

1998-01-01

This document presents the results of an engineering evaluation/cost analysis (EE/CA) that was conducted to evaluate alternatives to address final disposition of the 105-DR and 105-F Reactor Buildings (subsequently referred to as facilities), including the fuel storage basins (FSB) and below-grade portions of the reactors, excluding the reactor blocks. The reactor blocks will remain in a safe storage mode for up to 75 years as identified in the Record the Decision (ROD) (58 FR 48509) for the Environmental Impact Statement (EIS), Decommissioning of Eight Surplus Production Reactors at the Hanford Site, Richland, Washington (DOE 1992a). This EE/CA also addresses final disposition of four ancillary facilities: 116-D and 116-DR Exhaust Air Stacks, 117-DR Exhaust Filter Building, and 119-DR Exhaust Air Sample Building. The 105-DR and 105-F facilities are located in the 100-D and 100-F Areas of the Hanford Site. In November 1989, the 100 Area of the Hanford Site was placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The 100 Area NPL includes the 100-D Area (which includes the 100-DR site) and the 100-F Area, which are in various stages of the remediation process. It has been determined by DOE that hazardous substances in the 105-DR, 105-F, and the four ancillary facilities may present a potential threat to human health or the environment, and that a non-time critical removal action at these facilities is warranted. To help determine the most appropriate action, DOE, in cooperation with the Washington State Department of Ecology (Ecology) and the EPA, has prepared this EE/CA. The scope of the evaluation includes the 105-DR and 105-F facilities and the four ancillary facilities. The 116-DR and 117-DR facilities are located within the boundaries of the 105-DR Large Sodium Fire Facility Treatment, Storage, and Disposal (TSD) unit, which is

Design Guide for Category I reactors critical facilities

International Nuclear Information System (INIS)

Brynda, W.J.; Powell, R.W.

1978-08-01

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

Annual report to the Laser Facility Committee 1986

International Nuclear Information System (INIS)

1986-01-01

This paper is the annual report of the Science and Engineering Research Council, research and development work carried out at the Central Laser Facility, Rutherford Laboratory, United Kingdom, 1985/6. Part I contains the technical details of the studies of the High Power Laser scientific programme and Laser Support Facility, as well as the Laser Research and Development investigations. Part II concerns the application of UV lasers to microcircuit fabrication. (UK)

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

International Nuclear Information System (INIS)

1997-01-01

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

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.

Research nuclear reactor RA - Annual Report 1991

International Nuclear Information System (INIS)

Sotic, O.

1992-01-01

Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Construction of some support elements is almost finished by the local staff. The Institute has undertaken this activity in order to speed up the ending of the project. If all the planned instrumentation would not arrive until the end of March 1992, it would not be possible to start the RA reactor testing operation in the first part of 1993, as previously planned. In 1991, 53 staff members took part in the activities during 1991, which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

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

International Nuclear Information System (INIS)

Bissani, M; O'Kelly, D S

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Bissani, M; O' Kelly, D S

2006-05-08

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

Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

International Nuclear Information System (INIS)

Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

1995-01-01

A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

Reactor feedwater facility

Energy Technology Data Exchange (ETDEWEB)

Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

1996-04-30

In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.

Operation and maintenance of the RB reactor, Annual report for 1976

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1976-01-01

Due to its flexibility and relatively simple construction the RB reactor enabled direct measurements of a series of physical parameters. During 1976 the reactor operation exceeded the plan due to preparation of special experiments planned for the next period. It is planned to operate the reactor at higher power levels (50 W - 10 kw). A need for increasing the neutron flux a neutron converter was built in 1976. preliminary measurements showed that placing the neutron converter next to the reactor vessel enables achievement of irradiation and dosimetry measurements in the fast neutron flux. It is planned to purchase highly enriched fuel for the neutron converter. This annual report includes 5 Annexes with data concerning: operation, irradiation field around the RB reactor, maintenance of reactor components and instrumentation, purchase of new equipment, and the program for training reactor operators

The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

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

N Reactor Deactivation Program Plan

International Nuclear Information System (INIS)

Walsh, J.L.

1993-12-01

This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities · in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually

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

Energy Technology Data Exchange (ETDEWEB)

Middleton, A; Law, G C [General Atomic Co., San Diego, CA (United States)

1974-07-01

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 10{sup 14} n/cm{sup 2}-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 10{sup 17} n/cm{sup 2}-sec. The pulse width at

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

International Nuclear Information System (INIS)

Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

2009-01-01

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

The advanced neutron source - A world-class research reactor facility

International Nuclear Information System (INIS)

Thompson, P.B.; Meek, W.E.

1993-01-01

The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

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

International Nuclear Information System (INIS)

Vernetson, W.G.

1991-01-01

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

Hanford Facility Annual Dangerous Waste Report Calendar Year 2002

International Nuclear Information System (INIS)

FR-EEMAN, D.A.

2003-01-01

Hanford CY 2002 dangerous waste generation and management forms. The Hanford Facility Annual Dangerous Waste Report (ADWR) is prepared to meet the requirements of Washington Administrative Code Sections 173-303-220, Generator Reporting, and 173-303-390, Facility Reporting. In addition, the ADWR is required to meet Hanford Facility RCRA Permit Condition I.E.22, Annual Reporting. The ADWR provides summary information on dangerous waste generation and management activities for the Calendar Year for the Hanford Facility EPA ID number assigned to the Department of Energy for RCRA regulated waste, as well as Washington State only designated waste and radioactive mixed waste. The Solid Waste Information and Tracking System (SWITS) database is utilized to collect and compile the large array of data needed for preparation of this report. Information includes details of waste generated on the Hanford Facility, waste generated offsite and sent to Hanford for management, and other waste management activities conducted at Hanford, including treatment, storage, and disposal. Report details consist of waste descriptions and weights, waste codes and designations, and waste handling codes. In addition, for waste shipped to Hanford for treatment and/or disposal, information on manifest numbers, the waste transporter, the waste receiving facility, and the original waste generators are included. In addition to paper copies, electronic copies of the report are also transmitted to the regulatory agency

Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

International Nuclear Information System (INIS)

1988-11-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

Research nuclear reactor RA, Annual Report 2001

International Nuclear Information System (INIS)

Sotic, O.

2002-01-01

During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001 [sr

Environmental assessment for the deactivation of the N Reactor facilities

International Nuclear Information System (INIS)

1995-05-01

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)

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

International Nuclear Information System (INIS)

Langton, C.A.; Stefanko, D.B.; Serrato, M.G.; Blankenship, J.K.; Griffin, W.G.; Long, J.T.

2013-01-01

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

Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

International Nuclear Information System (INIS)

Beller, D.E.; Ott, K.O.; Terry, W.K.

1987-01-01

A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

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

Design of Safety Parameter Monitoring Function in a Research Reactor Facility

International Nuclear Information System (INIS)

Park, Jaekwan; Suh, Yongsuk

2014-01-01

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

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

International Nuclear Information System (INIS)

Chilian, C.; Kennedy, G.

2010-01-01

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)

TIT reactor laboratory course using JAERI and PNC large experimental facilities

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Obara, Toru; Ohtani, Nobuo.

1995-01-01

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)

WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

International Nuclear Information System (INIS)

Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

1982-01-01

At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

International Nuclear Information System (INIS)

1989-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

International Nuclear Information System (INIS)

Thomas, M.L.

1998-02-01

This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem) 2 which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem)

Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

Energy Technology Data Exchange (ETDEWEB)

Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

1998-02-01

This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem){sup 2} which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem).

Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

International Nuclear Information System (INIS)

Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

1975-01-01

A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

Experimental Facilities for Performance Evaluation of Fast Reactor Components

International Nuclear Information System (INIS)

Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

2013-01-01

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

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

International Nuclear Information System (INIS)

2002-01-01

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

Neutron beam facilities at the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Kim, S.

2003-01-01

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

Annual progress report 1981

International Nuclear Information System (INIS)

1982-01-01

This annual progress report of the CEA Protection and Nuclear Safety Institut outlines a brief description of the progress made in each section of the Institut. Research activities of the Protection department include, radiation effects on man, radioecology and environment radioprotection techniques. Research activities of the Nuclear Safety department include, reactor safety analysis, fuel cycle facilities safety analysis, safety research programs. The third section deals with nuclear material security including security of facilities, security of nuclear material transport and monitoring of nuclear material management [fr

Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

International Nuclear Information System (INIS)

Chatt, A.; Holzbecher, J.

1990-01-01

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

Operation of the High Flux Reactor. Annual report 1985

International Nuclear Information System (INIS)

1985-01-01

This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

International Nuclear Information System (INIS)

1991-09-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

International Nuclear Information System (INIS)

1992-08-01

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

Education and training at the Rensselaer Polytechnic Institute reactor critical facility

International Nuclear Information System (INIS)

Harris, D.R.

1989-01-01

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 UO 2 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

Liquid Metal Fast Breeder Reactor Program: Argonne facilities

International Nuclear Information System (INIS)

Stephens, S.V.

1976-09-01

The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. 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 appended

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

Energy Technology Data Exchange (ETDEWEB)

Lee, A G; Lidstone, R F; Bishop, W E; Talbot, E F; McIlwain, H [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

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 MW{sub t}, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 10{sup 18} n{center_dot}m{sup -2}{center_dot}s{sup -1} in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs.

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

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 (U 3 Si 2 ) 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 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C 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)

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

International Nuclear Information System (INIS)

1982-01-01

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

The application of SQUG to non-reactor facilities

International Nuclear Information System (INIS)

Hoskins, R.S.

1993-01-01

DOE Order 6430.1A mandates that facilities be designed and constructed to withstand Natural Phenomena Hazards in accordance with their hazard level. DOE has a program in progress to evaluate and then upgrade many of their existing medium and high hazard facilities where release of hazardous materials to the environment is a concern. This paper addresses a useful methodology which has been applied by SRS to evaluate and qualify equipment to withstand the ravenousness of earthquakes. The Seismic Qualification Utility Group was formed by a group of Electric Power Utilities whose Nuclear Power Plants predated the 10CFR50 Environmental design requirements to develop a methodology to evaluate and upgrade their operating plants against seismic events in answer to NRC generic letter USI-A46. SRS participated in this organization, since it operated reactors designed and constructed in the 1950's, and the application of the SQUG methodology was obvious. Nuclear Material Processing and Handling (NMPH) facilities utilize equipment similar to the nuclear industry, in fact, to industry in general. Consequently, it made sense to apply SQUG methodology to evaluate and qualify equipment in NMPH facilities against earthquakes. In order to utilize SQUG methodology, some changes are required since the goal of safe shutdown of a NMPH facility differs from a nuclear reactor, consequently, the generation of a Safe Shutdown Equipment List is modified by the requirements of each particular facility. Once the Safety Class Equipment List is developed denoting the qualification requirement for each piece of equipment, the SQUG walk downs can be performed as they are in commercial nuclear plants. SQUG methodology offers a cost effective approach for the seismic qualification of equipment at existing DOE NMPH facilities. SQUG cannot be applied to new or unique equipment since the experience database doesn't contain the needed information

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

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

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

The advanced test reactor national scientific user facility advancing nuclear technology

International Nuclear Information System (INIS)

Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

2009-01-01

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

Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

International Nuclear Information System (INIS)

Klueh, R.L.; Bloom, E.E.

1981-12-01

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

Nuclear reactors built, being built, or planned: 1989

International Nuclear Information System (INIS)

1990-06-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1989. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. Information is presented in five parts, each of which is categorized by primary function or purpose: civilian, production, military, export, and critical assembly facilities

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

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

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

1996 environmental monitoring report for the Naval Reactors Facility

International Nuclear Information System (INIS)

1996-01-01

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)

Annual report ''nuclear safety in France''; Le rapport annuel ''la surete nucleaire en France''

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This document is the 2001 annual report of the French authority of nuclear safety (ASN). It summarizes the highlights of the year 2000 and details the following aspects: the nuclear safety in France, the organization of the control of nuclear safety, the regulation relative to basic nuclear facilities, the control of facilities, the information of the public, the international relations, the organisation of emergencies, the radiation protection, the transport of radioactive materials, the radioactive wastes, the PWR reactors, the experimental reactors and other laboratories and facilities, the nuclear fuel cycle facilities, and the shutdown and dismantling of nuclear facilities. (J.S.)

SERC Central Laser Facility annual report 1992

International Nuclear Information System (INIS)

1992-01-01

In this 1992 Annual Report to the Laser Facility Committee of the Science and Engineering Research Council, the Central Laser Facility at Rutherford Appleton Laboratory, technical progress is described and mid-term organizational goals outlined. Outstanding among recent achievements is the work on plasma heating being undertaken on the Sprite facility using the ultra-bright KrF laser pumped Raman beams. Two-beam operation at power levels approaching 2 TW in 10 ps are hoped for. On a four year timescale the Titania system will provide four Raman beams of exceptional brightness and power up to 20TW in 10ps. The other high power laser facility, Vulcan is also producing exciting work. Progress in nanosecond studies using Raman spectroscopy have produced the first Raman spectrum of solvated Buckmister fullerene and direct observation of the separation of germinate ion pairs, as well as information on the behaviour of a single base in an oligonuclide chain. Phase boundaries for the solidification of a two dimensional electron fluid have been determined in a Gallium Arsenide heterojunction. Despite staff number attrition, operation and development of the facilities have continued successfully. (UK)

Neutron beam facilities at the Australian Replacement Research Reactor

International Nuclear Information System (INIS)

Kennedy, Shane; Robinson, Robert; Hunter, Brett

2001-01-01

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 10 14 n/cm 2 /sec and a liquid D 2 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)

Construction of the neutron beam facility at Australia's OPAL research reactor

International Nuclear Information System (INIS)

Kennedy, Shane J.

2006-01-01

Australia's new research reactor, OPAL, has been designed principally for neutron beam science and radioisotope production. It has a capacity for 18 neutron beam instruments, located at the reactor face and in a neutron guide hall. The neutron beam facility features a 20 l liquid deuterium cold neutron source and cold and thermal supermirror neutron guides. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, when criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. This paper will outline the key features of the OPAL reactor, and will describe the neutron beam facility in particular. The status of the construction and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed

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

International Nuclear Information System (INIS)

Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

1986-01-01

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.)

Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1985-03-01

This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

International Nuclear Information System (INIS)

Hardt, P. von der; Roettger, H.

1981-01-01

Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

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.

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

International Nuclear Information System (INIS)

Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

1983-02-01

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

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

International Nuclear Information System (INIS)

Montgomery, B.H.; Thoms, K.R.; West, C.D.

1987-09-01

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

Use Of Cementitious Materials For SRS Reactor Facility In-Situ Decommissioning - 11620

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

Annual report on the state of RB reactor components and equipment, december 1999

International Nuclear Information System (INIS)

Milosevic, M.

1999-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 13 times, meaning that experiments were done with 13 configurations of the reactor core. Total reactor operation amounted to 84 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1999, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

Annual report on the state of RB reactor components and equipment, december 1998

International Nuclear Information System (INIS)

Milosevic, M.

1998-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

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

International Nuclear Information System (INIS)

1997-01-01

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 41 Ar) were well within regulatory limits

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.

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

International Nuclear Information System (INIS)

Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

1983-01-01

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.) [de

Proposal for a seismic facility for reactor safety research

International Nuclear Information System (INIS)

Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

1976-07-01

Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

International Nuclear Information System (INIS)

Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

1992-07-01

This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

Annual report in compliance with the reactor sharing program, September 1, 1994--August 31, 1995

International Nuclear Information System (INIS)

Karam, R.A.

1997-01-01

This report contains information with regard to facilities utilization, descriptions (brief), personnel, organization, and programs of the Neely Nuclear Research Center (NNRC) at the Georgia Institute of Technology. The NNRC has two major facilities: the Georgia Tech Research Reactor and the Hot Cell Laboratory. This report of NNRC utilization is prepared in compliance with the contract requirements between the U.S. Department of Energy and the Georgia Institute of Technology. The NNRC is a participant in the University Reactor Sharing Program; as such, it makes available its 5 MW research reactor, its Co-60 irradiation facility and its activation analysis laboratory to large numbers of students and faculty from many universities and colleges

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

The Advanced Test Reactor Irradiation Facilities and Capabilities

International Nuclear Information System (INIS)

S. Blaine Grover; Raymond V. Furstenau

2007-01-01

The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Ishihara, Noriyuki

1985-01-01

The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

International Nuclear Information System (INIS)

Gieseler, Walter

1970-01-01

The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

Cold neutron PGAA facility developments at university research reactors in the USA

International Nuclear Information System (INIS)

Uenlue, K.; Rios-Martinez, C.

2005-01-01

The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

Situation of the radioactive waste management and the employee radiation exposure in commercial power generation reactor facilities in fiscal 1980

International Nuclear Information System (INIS)

1981-01-01

(1) Situation of the radioactive waste management in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the target dose around the sites by law in the radioactive waste management. The release of radioactive gaseous and liquid wastes and the storage of radioactive solid wastes in respective reactor facilities in fiscal 1980 are presented in tables (for the former, the data since 1971 are also given). The release control values were satisfied in all the facilities. (2) Situation of employe radiation exposure in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the permissible exposure doses by law. The Employe exposure doses in respective reactor facilities in fiscal 1980 are given in tables. All exposure doses were below the permissible levels. (J.P.N.)

A neutron radiography facility on the IRT-2000 reactor

International Nuclear Information System (INIS)

Khadduri, I.Y.

1976-01-01

A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges - 15066

International Nuclear Information System (INIS)

Sabharwall, P.; O'Brien, J.E.; Yoon, S.J.; Sun, X.

2015-01-01

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 3 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 circuits 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 Integrated 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. degrees), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF 4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼ 450 C. degrees. 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

Analysis of Opportunity to Create Self-Regulating Reactor Facility of Extra-Low Power

International Nuclear Information System (INIS)

Kazansky, Y.A.; Levtchenko, V.A.; Yuriev, Y.S.

2002-01-01

This paper deals with fundamental possibilities (economy, safety, self-regulation) of creating an extra-low power reactor facility for heat supply. It contains the results of calculations for thermal and fast neutron reactors. The concept of this type of a reactor had been developed by the contributors earlier

China institute of atomic energy annual report (1996)

International Nuclear Information System (INIS)

1997-01-01

The major achievements on scientific researches and production of CIAE during the year of 1996 was summarized in this annual report. It includes nuclear physics, high power laser, electro-physics reactor science and technology, radiochemistry. radioactive wastes treatment and disposal, analytical chemistry and nuclear safeguard techniques, isotopes, application of nuclear technique, radiation protection and environmental protection and others. New progress was made in proposing the China Experimental Fast Reactor (CEFR), China advanced research reactor (CARR) and Beijing radioactive nuclear Beam facility (BRNBF) which are the three major projects of CIAE

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

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

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

Nuclear reactors built, being built, or planned: 1986

International Nuclear Information System (INIS)

Carter, E.P.

1987-03-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1986, which are capable of sustaining a nuclear chain reaction. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commisssion; from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; and from US embassies of foreign countries. Information is presented in five parts, each of which is categorized by primary function or purpose: civilian, production, military, export, and critical assembly facilities

An independent safety assessment of Department of Energy nuclear reactor facilities: Procedures, operations and maintenance

International Nuclear Information System (INIS)

Toto, G.; Lindgren, A.J.

1981-02-01

The 1979 accident at the Three Mile Island commercial nuclear power plant has led to a number of studies of nuclear reactors, in both the public and private sectors. One of these is that of the Department of Energy's (DOE) Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, which has outlined tasks for assessment of 13 reactors owned by DOE and operated by contractors. This report covers one of the tasks, the assessment of procedures, operations, and maintenance at the DOE reactor facilities, based on a review of actual documents used at the reactor sites

Physics design of fast reactor safety test facilities for in-pile experiments

International Nuclear Information System (INIS)

Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

1976-01-01

A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

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

Energy Technology Data Exchange (ETDEWEB)

Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

2012-03-15

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)

THE WHITE SANDS MISSILE RANGE PULSED REACTOR FACILITY, MAY 1963

Energy Technology Data Exchange (ETDEWEB)

Long, Robert L.; Boor, R. A.; Cole, W. M.; Elder, G. E.

1963-05-15

A brief statement of the mission of the White Sands Missile Range Nuclear Effects Laboratory is given. The new Nuclear Effects Laboratory Facility is described. This facility consists of two buildings-a laboratory and a reactor building. The White Sands Missile Range bare critical assembly, designated as the MoLLY-G, is described. The MoLLY-G, an unreflected, unmoderated right circular cylinder of uranium-molybdenum alloy designed for pulsed operation, will have a maximum burst capability of approximately 2 x 10/sup 17/ fissions with a burst width of 50 microseconds. The reactor construction and operating procedures are described. As designed, the MoLLY-G will provide an intense source of pulsed neutron and gamma radiation for a great variety of experimental and test arrangements. (auth)

Safety report content and development for test loop facility on MARIA reactor

International Nuclear Information System (INIS)

Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

1982-01-01

A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

Annual report on the state of RB reactor components and equipment, December 1997

International Nuclear Information System (INIS)

Milosevic, M.

1997-12-01

According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1997 the reactor lattice was not changed due to application of the coupled fast-thermal core HERBE. Total reactor operation amounted to 69.5 Wh with 66 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment, plan for forming new HERBE core, plan for regular annual maintenance of the reactor, and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

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

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

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

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)

1996-01-01

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.)

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

Energy Technology Data Exchange (ETDEWEB)

None

1965-06-01

This report 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 procedures are presented in another report.

Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

International Nuclear Information System (INIS)

Sarkulov, M.K.

2015-01-01

There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

Research reactor facilities, recent developments at Imperial College, London

International Nuclear Information System (INIS)

Franklin, S.J.; Goddard, A.J.H.; O Connell, J.

1998-01-01

The 100 kW CONSORT pool-type reactor is now the only Research Reactor in the UK. Because of its strategic importance, Imperial College is continuing and accelerating a programme of refurbishment of the control system, and planning for a further fuel charge. These plans are described and the progress to date discussed. To this end, a description of the enhanced Safety Case being written is provided here and refueling plans discussed. The current range of facilities available is described, and future plans highlighted. (author)

NRX and NRU reactor research facilities and irradiation and examination charges

International Nuclear Information System (INIS)

1960-08-01

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.

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

International Nuclear Information System (INIS)

Weightman, M.W.; Willby, C.R.

1990-01-01

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

A tokamak reactor with servicing capability

International Nuclear Information System (INIS)

Mitchell, J.T.D.; Hollis, A.

1976-01-01

A conceptual design for a Tokamak reactor with practical facilities for the regular replacement of blanket components after the inevitable damage from neutron irradiation, and fatigue is described. This essential facility has been largely ignored in published fusion reactor designs. One exception is the inertially-confined Saturn proposal. Tokamak and other toroidal closed-line systems have very complex geometries and sub-system requirements, which result in blanket servicing being a very difficult problem. In the concept described the magnet shield is divided into two structures - an outer permanent one with access doors and an inner shield, part of and supporting the blanket inside. Servicing access is horizontally between the toroidal magnet coils, after moving some outer poloidal magnet coils. The reactor, reactor hall, workshops and remote-handling facilities are described, and the servicing requirements discussed. The important servicing operation is the remote replacement of radiation damaged blanket and shield - divided in this design into 20 sectors, each weighing 75-100 tons and 11-12 metres high. Analysis of the operation indicates that if one sector can be replaced during a single weekend - i.e. a period of low power demand - then the annual reactor-generator availability allowing as well for the general plant servicing should be >0.9. This level of availability should meet the requirements of generating authorities but the facilities, equipment and workshops necessary may be complex and expensive

An independent safety assessment of Department of Energy nuclear reactor facilities: Safety overview and management function

International Nuclear Information System (INIS)

Booth, M.; Brodsky, R.S.; Frankhouser, W.L.

1981-02-01

The Under Secretary of Energy established the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee in October, 1979, in the aftermath of the Three Mile Island (TMI) nuclear accident, to assess the adequacy of training of personnel at DOE nuclear facilities. Subsequently, in February, 1980, the charge to this Committee was modified to assess all implications of the Kemeny Commission report on TMI with regard to DOE nuclear reactors, excluding those in the Division of Naval Reactors. The modified charge was also limited, for the time being, to reactor facilities instead of all nuclear facilities. This report describes the portion of the revised assessment activities that was assigned to the Assessment Support Team

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

1975-12-01

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

Abnormal Events for Reactor System and Facilities in HANARO

Energy Technology Data Exchange (ETDEWEB)

Choi, Ho Young; Lee, B. H.; Lee, M.; Kang, I. H.; Lee, U. G.; Sin, H. C.; Park, C. Y.; Song, B. S.; Lee, S. H.; Han, J. S

2006-12-15

This report gathers abnormal events related to reactor system and facilities of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

International Nuclear Information System (INIS)

Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

2009-03-01

The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

Domestic safeguards: annual report to Congress, fiscal year 1978

International Nuclear Information System (INIS)

1979-01-01

The annual report includes an assessment of the effectiveness and adequacy of safeguards at facilities and activities licensed by the Commission. The report details NRC's criteria for judging the adequacy of safeguards at fuel cycle facilities; the report also summarizes actions required by NRC at any fuel facility whose safeguards systems are judged to provide less than high assurance protection against our design threat. The report also contains a discussion of NRC's criteria for safeguards adequacy at nuclear reactors and for transportation activities

Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

International Nuclear Information System (INIS)

Ahnstroem, G.; Burtscher, A.; Casta, J.

1967-01-01

An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

The reactor and cold neutron research facility at NIST

Energy Technology Data Exchange (ETDEWEB)

Prask, H J; Rowe, J M [Reactor Radiation Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

1992-07-01

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 D{sub 2}O/H{sub 2}O ice cold source. (author)

The reactor and cold neutron research facility at NIST

International Nuclear Information System (INIS)

Prask, H.J.; Rowe, J.M.

1992-01-01

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 D 2 O/H 2 O ice cold source. (author)

Status of national programmes on fast reactors 1998/99. 32nd annual meeting. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-07-01

Over the past 32 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1998/1999, as reported at the 32. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States.

Status of national programmes on fast reactors 1998/99. 32nd annual meeting. Working material

International Nuclear Information System (INIS)

1999-01-01

Over the past 32 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1998/1999, as reported at the 32. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

Energy Technology Data Exchange (ETDEWEB)

Ramirez Aviles, Camila A. [ORNL; Rao, Nageswara S. [ORNL

2018-01-01

We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventional majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.

Annual report of Nuclear Engineering Research Laboratory, University of Tokyo in fiscal 1990

International Nuclear Information System (INIS)

1991-01-01

In this annual report, the activities of research and education and the state of operation of the research facilities in this Laboratory in fiscal year 1990 are summarized. There are four large research facilities in this Laboratory, that is, the fast neutron source reactor 'Yayoi', the electron beam linear accelerator, the nuclear fusion reactor blanket experiment device and the heavy ion irradiation research facility. Those are used to execute research and education in the wide fields of atomic energy engineering, and put to the common utilization by universities in whole Japan. The results of the research with these facilities have been reported in the separate reports. The research aims at developing the most advanced and new fields in nuclear reactor engineering, and includes the engineering of the first wall and the fuel cycle for nuclear fusion reactors, electromagnetic structure engineering, AI and robotics, quantum beam engineering, the design of new type reactors, the basic process of radiochemistry and so on. The report on the course of the large scale facilities, research activities, the publication of research, education and the events in the Laboratory in the year are described. (K.I.)

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

International Nuclear Information System (INIS)

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

Dual Axis Radiographic Hydrodynamic Test Facility mitigation action plan. Annual report for 1998

International Nuclear Information System (INIS)

Haagenstad, T.

1999-01-01

This Mitigation Action Plan Annual Report (MAPAR) has been prepared as part of implementing the Dual Axis Radiographic Hydrodynamic Test Facility (DARHT) Mitigation Action Plan (MAP) to protect workers, soils, water, and biotic and cultural resources in and around the facility

Dual Axis Radiographic Hydrodynamic Test Facility mitigation action plan. Annual report for 1998

Energy Technology Data Exchange (ETDEWEB)

Haagenstad, T.

1999-01-15

This Mitigation Action Plan Annual Report (MAPAR) has been prepared as part of implementing the Dual Axis Radiographic Hydrodynamic Test Facility (DARHT) Mitigation Action Plan (MAP) to protect workers, soils, water, and biotic and cultural resources in and around the facility.

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

International Nuclear Information System (INIS)

Gramajo, M.; García, J.; Marcel, C.P.

2013-01-01

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 D 2 O 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)

Progress on ANSTO'S OPAL reactor project and its future importance as the centrepiece of ANSTO'S facilities

International Nuclear Information System (INIS)

Smith, I.O.

2006-01-01

Full text: After an intensive process of analysis, the Australian government approved the construction of a multi-purpose research reactor in 1997. Following the conduct of a comprehensive tender evaluation process in 1998-2000, INVAP was contracted to construct a 20 MW open pool research reactor and associated neutron beam facilities. The construction of the reactor is now almost complete, and we have commenced cold commissioning. ANSTO has applied for an operating licence, and we hope for a decision on that application in June, following the consideration by the regulator of the results of cold commissioning. The OPAL reactor will provide neutrons to a world-class neutron beam facility, in which a number of the instruments will have the best performance available in the world to date. We intend to establish the Bragg Institute as a regional centre of excellence on neutron beam science, with a significant number of international scientists using the facility to produce cutting edge science in the fields of biology, materials science, food science and other area. The reactor also has extensive irradiation facilities within the reflector vessel. These facilities will be used to produce medical isotopes - ANSTO supplies the bulk of the Australian market and also exports into this region - and for the transmutation doping of silicon ingots for semiconductor manufacture. There are also a number of pneumatically loaded radiation facilities allowing for short term irradiation of samples for such activities as neutron activation analysis

Annual meeting on nuclear technology 2005. Proceedings

International Nuclear Information System (INIS)

2005-03-01

The proceedings of the annual meeting on nuclear technology 2005 covers the following issues: (1) reactor physics and methods of calculation: design and transients; method development and validation; (2): thermodynamics and fluid dynamics: analytical thermohydraulics for existing reactors; experiments and operational behavior; analytical methods for innovative reactors; (3) Safety of nuclear installations - methods, analysis, results: special problems; PSA and in-vessel phenomena; ex-vessel phenomena; (4) front end and back end of the fuel cycle, radioactive waste, storage: intermediate storage of fuel elements, waste treatment, (5) fuel elements and core components: fuel elements, new methods in the interpretation, manufacturing and service; (6) operation of nuclear installations: experience with the operation of NPPs; management systems, digital instrumentation and control of NPPs revision management; (7) decommissioning of nuclear installations: concepts and strategies for decommissioning and dismantling; experiences with decommissioning projects; (8) fusion technology: fusion facilities; materials and test facility; cryo technique and simulations; (9) research reactors: building new and backfitting of existing research reactors; current development; dismantling of research reactors; (10) advanced reactor concepts, energy systems, energy economics; (11) communication with the public; (12) component materials, fabrication and service behavior: degradation effects of component materials; component behavior; (13): radiation protection: PSA and in-vessel phenomena, ex-vessel phenomena.

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

International Nuclear Information System (INIS)

Alloni, L.; Venturelli, A.; Meloni, S.

1990-01-01

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.)

Modernization of the facilities of the TRIGA Mark III reactor of ININ

International Nuclear Information System (INIS)

Mendez T, D.; Flores C, J.

2016-09-01

The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

Annual Summary of the Integrated Disposal Facility Performance Assessment 2012

Energy Technology Data Exchange (ETDEWEB)

Khaleel, R. [INTERA, Austin, TX (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

2012-12-27

An annual summary of the adequacy of the Hanford Immobilized Low-Activity Waste (ILAW) Performance Assessment (PA) is required each year (DOE O 435.1 Chg 1,1 DOE M 435.1-1 Chg 1;2 and DOE/ORP-2000-013). The most recently approved PA is DOE/ORP-2000-24.4 The ILAW PA evaluated the adequacy of the ILAW disposal facility, now referred to as the Integrated Disposal Facility (IDF), for the safe disposal of vitrified Hanford Site tank waste.

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

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

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.)

Material accountancy and control practice at a research reactor facility

International Nuclear Information System (INIS)

Bouchard, J.; Maurel, J.J.; Tromeur, Y.

1982-01-01

This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

Annual report 90. Institute for advanced materials

International Nuclear Information System (INIS)

1991-01-01

The Annual Report 1990 of the Institute for Advanced Materials of the JRC highlights the Scientific Technical Achievements and presents in the Annex the Institute's Competence and Facilities available to industry for services and research under contract. The Institute executed in 1990 the R and D programme on advanced materials of the JRC and contributed to the programmes: reactor safety, radio-active waste management, fusion technology and safety, nuclear fuel and actinide research. The supplementary programme: Operation of the High Flux Reactor is presented in condensed form. A full report is published separately

Facility effluent monitoring plan for the 300 Area Fuels Fabrication Facility

International Nuclear Information System (INIS)

Nickels, J.M.; Brendel, D.F.

1991-11-01

A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring system by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The Fuel Fabrication Facility in the Hanford 300 Area supported the production reactors from the 1940's until they were shut down in 1987. Prior to 1987 the Fuel Fabrication Facility released both airborne and liquid radioactive effluents. In January 1987 the emission of airborne radioactive effluents ceased with the shutdown of the fuels facility. The release of liquid radioactive effluents have continued although decreasing significantly from 1987 to 1990

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

International Nuclear Information System (INIS)

2013-01-01

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

Lining facility for FBR type reactor

International Nuclear Information System (INIS)

Shimano, Kunio.

1991-01-01

In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

International Nuclear Information System (INIS)

Hine, R.E.

1980-09-01

This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

The Advanced Neutron Source (ANS) project: A world-class research reactor facility

International Nuclear Information System (INIS)

Thompson, P.B.; Meek, W.E.

1993-01-01

This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5x10 19 m -2 ·sec -1 . Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities

RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1987-01-01

RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

Annual Report to Congress of the Atomic Energy Commission for 1966

Energy Technology Data Exchange (ETDEWEB)

Seaborg, Glenn T.

1967-01-31

The document represents the 1966 Annual Report of the Atomic Energy Commission (AEC) to Congress. The report opens with a Summary of Atomic Energy Programs in 1966 and includes 18 Chapters, 11 appendices and an index. Chapters are as follows: (1) Licensing and Regulating the Atom; (2) Reactor and Other Nuclear Facility Licensing; (3) The Regulation of Radioactive Materials; (4) Source and Special Nuclear Materials; (5) The Nuclear Defense Effort; (6) Naval Propulsion Reactors; (7) Reactor Development and Technology; (8) Space Nuclear Systems; (9) Isotopic Heat and Power Applications; (10) Isotopic Radiation Applications; (11) The Plowshare Program; (12) International Cooperation Activities; (13) Research Facilities and Projects; (14) Nuclear Education and Training; (15) Informational Activities; (16) Operational Safety; (17) Industrial Participation Aspects; and, (18) Administrative and Management Matters.

Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

International Nuclear Information System (INIS)

Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

1997-01-01

Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

Construction of the Neutron Beam Facility at Australia's OPAL Research Reactor

International Nuclear Information System (INIS)

Kennedy, J.S.

2005-01-01

Full text: Australia's new research reactor, OPAL, has been designed for high quality neutron beam science and radioisotope production. It has a capacity for eighteen neutron beam instruments to be located at the reactor face and in a neutron guide hall. The new neutron beam facility features a 20 litre liquid deuterium cold neutron source and supermirror neutron reflecting guides for intense cold and thermal neutron beams. Nine neutron beam instruments are under development, of which seven are scheduled for completion in early 2007. The project is approaching the hot-commissioning stage, where criticality will be demonstrated. Installation of the neutron beam transport system and neutron beam instruments in the neutron guide hall and at the reactor face is underway, and the path to completion of this project is relatively clear. The lecture will outline Australia's aspirations for neutron science at the OPAL reactor, and describe the neutron beam facility under construction. The status of this project and a forecast of the program to completion, including commissioning and commencement of routine operation in 2007 will also be discussed. This project is the culmination of almost a decade of effort. We now eagerly anticipate catapulting Australia's neutron beam science capability to meet the best in the world today. (author)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1985-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

1982-09-01

Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

Project RA Research nuclear reactor - Annual report 1993 with comparative review for the period 1991 - 1993

International Nuclear Information System (INIS)

1993-12-01

Research reactor RA Annual report for year 1993 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. The ninth separate annex deals with the feasibility of RA reactor applications. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

Power reactors in member states

International Nuclear Information System (INIS)

1975-01-01

This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

Emergency planning and response: An independent safety assessment of Department of Energy nuclear reactor facilities

International Nuclear Information System (INIS)

Knuth, D.; Boyd, R.

1981-02-01

The Department of Energy (DOE) has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the recommendations contained in the President's Commission Report on the Three Mile Island (TMI) Accident (the Kemeny Commission report) that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors have been reviewed. The assessments of the 13 facilities are based on information provided by the individual operator organizations and/or cognizant DOE Field Offices. Additional clarifying information was supplied in some, but not all, instances. This report indicates how these 13 reactor facilities measure up in light of the Kemeny and other TMI-related studies and recommendations, particularly those that have resulted in upgraded Nuclear Regulatory Commission (NRC) requirements in the area of emergency planning and response

A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; An, Shigehiro.

1983-01-01

Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

Scottish Universities Research and Reactor Centre annual report 1987-1988

International Nuclear Information System (INIS)

Whitley, J.E.

1988-01-01

The Scottish Universities Research and Reactor Centre (SURRC) provides facilities for research in isotopic, nuclear and earth sciences and collaborates with Scottish University departments on a wide range of research topics. One of its main areas of work is the Isotope Geology Unit. This has worked with the Nuclear Medicine Unit on the application of enriched stable isotope tracers in the biological and clinical sciences. The measurement of radioactive isomers is applied to quaternary geology, archaeology, nuclear medicine, health physics, oceanography, atomospheric sciences, environmental chemistry, nuclear waste disposal and mathematical modelling of the environment. There are also radiocarbon dating facilities. The facilities and the research undertaken at the Centre in the year 1987-1988, the Centre's twenty-fifth year are summarized in this report. (U.K.)

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

Energy Technology Data Exchange (ETDEWEB)

1984-01-01

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.

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

International Nuclear Information System (INIS)

1984-01-01

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

Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-02-01

The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1978-10-01

Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

International Nuclear Information System (INIS)

Tanaka, Nobuo

2017-01-01

Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

Annual report of Nuclear Engineering Research Laboratory, University of Tokyo in fiscal 1992

International Nuclear Information System (INIS)

1993-07-01

In this annual report, the activities of education and research, the state of operation of research facilities and others in Nuclear Engineering Research Laboratory, University of Tokyo in fiscal year 1992 are summarized. In this Laboratory, there are four large research facilities, that are, the fast neutron source reactor 'Yayoi', the electron beam linac, the nuclear fusion reactor blanket experiment facility and the heavy irradiation research facility. Those are used for carrying out education and research in the wide fields of nuclear engineering, and are offered also for joint utilization. The results of research by using respective research facilities have been summarized in separate reports. The course of the management and operation of each research facility is described, and the research activities, the theses for doctorate and graduation these of teachers, personnel and graduate students in the Laboratory are summarized. (J.P.N.)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

1977-09-01

Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

1984-08-01

Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1976-09-01

Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

UCN-VCN facility and experiments in Kyoto University Reactor

International Nuclear Information System (INIS)

Kawabata, Yuji; Okumura, Kiyoshi; Utsuro, Masahiko

1993-01-01

An ultracold and very cold neutron facility was installed in Kyoto University Reactor (KUR). The facility consists of a very cold neutron (VCN) guide tube, a VCN bender, a supermirror neutron turbine and experimental equipments with ultracold neutrons (UCN). The properties of each equipments are presented. UCN is generated by a supermirror neutron turbine combined with the cold neutron source operated with liquid deuterium, and the UCN output spectrum was measured by the time-of-flight method. A gravity analyzer for high resolution spectroscopy and a neutron bottle for decay experiments are now developing as the UCN research in KUR. (author)

Pilot-scale reactor activation facility at SRL

International Nuclear Information System (INIS)

Bowman, W.W.

1976-01-01

The Hydrogeocemical and Stream Sediment Reconnaissance portion of the National Uranium Resource Evaluation program requires an analytical technique for uranium and other elements. Based on an automated absolute activation analysis technique using 252 Cf, a pilt-scale facility installed in a production reactor has provided analyses for 2800 samples. Key features include: an automated sample transport system, a delayed neutron detector, two GeLi detectors, a loader, and an unloader, with all components controlled by a microprocessor; a dedicated PDP-9 computer and pulse height analyzer; and correlation and reduction of acquired data by a series of programs using an IBM 360/195 computer. The facility was calibrated with elemental and isotopic standards. Results of analyses of standard reference materials and operational detection limits for typical sediment samples are presented. Plans to increase sample throughput are discussed briefly

Nuclear reactors built, being built, or planned, 1988

International Nuclear Information System (INIS)

1989-08-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1988. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington Headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables. Section 2 includes nuclear reactors that are operating, being built, or planned. Section 3 includes reactors that have been shut down permanently or dismantled

Annual report on operation and management of hot laboratories and facilities. From April 1, 2006 to March 31, 2007

International Nuclear Information System (INIS)

2008-02-01

This is an annual report in a fiscal year 2006 that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), the Research Hot Laboratory (RHL) and the other research hot facilities in the Department of Hot laboratories and facilities. In RFEF, destructive examinations of BWR fuel rods and re-assembly were carried out as PIEs for a fuel assembly irradiated for 5 cycles in the Fukushima-2 Nuclear Power Station Unit-1. Mechanical property measurement of high burn-up fuel rods were performed as spent fuel integrity test for long term dry storage in order to formulate guidelines and technical criteria. In WASTEF, Slow Strain Rate Tests (SSRT) and Uni-axial Constant Load Tensile tests (UCLT) of in-core materials in pressurized high-temperature water condition, stress corrosion cracking tests for high-performance fuel cladding material and calorific value measurement of pulse irradiated fuel in NSRR were carried out. In RHL, equipment un-installations and decontamination were performed to lead cells according to the decommissioning plan. And modification of fuel storage room were started in order to utilize the facility for un-irradiated fuel storage after a fiscal year 2007. In addition, management of the other research hot facilities (No.1 Plutonium Laboratory, No.2 Research Laboratory, No.4 Research Laboratory, Analytical Chemistry Laboratory, Uranium Enrichment Laboratory, (Simulation Test for Environmental Radionuclide Migration (STEM), Clean Laboratory for Environmental Analysis and Research (CLEAR) and fuel storage) were carried out. (author)

Saint-Laurent-des-Eaux nuclear facilities. 2009 annual report

International Nuclear Information System (INIS)

2010-01-01

This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Saint-Laurent-des-Eaux, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

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

International Nuclear Information System (INIS)

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

Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

International Nuclear Information System (INIS)

Fujiki, K.

1992-01-01

The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

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

International Nuclear Information System (INIS)

Nystrand, A.C.; Lassing, A.

1999-01-01

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

2016 Annual Inspection and Radiological Survey Results for the Piqua, Ohio, Decommissioned Reactor Site, July 2016

Energy Technology Data Exchange (ETDEWEB)

Zimmerman, Brian [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

2016-07-01

This report presents the findings of the annual inspection and radiological survey of the Piqua, Ohio, Decommissioned Reactor Site (site). The decommissioned nuclear power demonstration facility was inspected and surveyed on April 15, 2016. The site, located on the east bank of the Great Miami River in Piqua, Ohio, was in fair physical condition. There is no requirement for a follow-up inspection, partly because City of Piqua (City) personnel participated in a March 2016 meeting to address reoccurring safety concerns. Radiological survey results from 104 locations revealed no removable contamination. One direct beta activity reading in a floor drain on the 56-foot level (1674 disintegrations per minute [dpm]/100 square centimeters [cm2]) exceeded the minimum detectable activity (MDA). Beta activity has been detected in the past at this floor drain. The reading was well below the action level of 5000 dpm/100 cm2.

Nuclear reactor built, being built, or planned

International Nuclear Information System (INIS)

1991-06-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

Reactor containing facility

International Nuclear Information System (INIS)

Akagawa, Katsuhiko.

1992-01-01

A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

The Annual Neutron School: Program and Facility for Nuclear Science and Technology

International Nuclear Information System (INIS)

Dingle, C.A.M.; Bautista, U.M.; Jecong, J.F.M.; Hila, F.C.; Astronomo, A.A.; Olivares, R.U.; Guillermo, N.R.D.; Ramo, M.E.S.K.V.; Saligan, P.P.

2015-01-01

The core realization of the mandate of the Philippine Nuclear Research Institute (PNRI) is the establishment and utilization of major nuclear facilities in lieu of the decommissioned research reactor. To address the need for manpower in the future, the applied physics research section (APRS) of the PNRI has initiated capacity building in the use and operation of small neutron sources which attempts to re-establish, develop and sustain expertise in nuclear science and technology. These activities have provided the theoretical and experimental training of young professionals and scientist of the institute which, consequently, resulted in the conceptualization of the Annual Neutron School (ANS).The ANS provides training and teaching environments for the young generation who will operate, utilize and regulate future nuclear facilities. More importantly, it demonstrates and presents the acquired knowledge and research outputs by the staff via “train a trainer” concept to an audience of junior undergraduate students. The successful implementation of the ANS has been participated by selected universities within Metro Manila and was able to train a number of students since its establishment in 2013. The program offers training, education, and R & D in the basic nuclear instrumentation and techniques which includes (1) characterization of different neutron sources – AmBe, PuBe and Cf-252; (2) development of Neutron Activation Analysis (NAA) technique using a portable neutron source for non-destructive elemental analysis; (3) utilization of MCNP (Monte Carlo N-Particle) code for verification of experimental data on neutron characterization, radiation dosimetry, detector design, calibration and efficiency and TRIGA fuel assembly configuration for sub-critical experiments. (author)

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-11-01

Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

Reactor Engineering Department annual report

International Nuclear Information System (INIS)

Matsuura, S.; Nakahara, Y.; Takano, H.

1983-09-01

Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

Nuclear reactors built, being built, or planned 1992

International Nuclear Information System (INIS)

1993-07-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1992. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. Information is presented on five parts: Civilian, Production, Military, Export and Critical Assembly

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

International Nuclear Information System (INIS)

Severin, V.P.

2007-01-01

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

Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

International Nuclear Information System (INIS)

Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

1990-11-01

The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

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

International Nuclear Information System (INIS)

2013-01-01

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

Reports and operational engineering: An independent safety assessment of Department of Energy nuclear reactor facilities

International Nuclear Information System (INIS)

Rochman, A.; Washburn, B.W.

1981-02-01

The Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, established via an October 24, 1979 memorandum from the Department of Energy (DOE) Under Secretary, was instructed to review the ''Kemeny Commission'' recommendations and to identify possible implications for DOE's nuclear facilities. As a result of this review, the Committee recommended that DOE carry out assessments in seven categories. The assessments would address specific topics identified for each category as delineated in the NFPQT ''Guidelines for Assessing the Safe Operation of DOE-Owned Reactors,'' dated May 7, 1980. The Committee recognized that similar assessments had been ongoing in the DOE program and safety overview organizations since the Three Mile Island nuclear accident and it was the Committee's intent to use the results of those ongoing assessments as an input to their evaluations. This information would be supplemented by additional studies consisting of the subject-related documents used at each reactor facility studied, and an on-site review of these reactor facilities by professional personnel within the Department of Energy, its operating contractors and independent consultants. 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

1982-09-01

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.

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

International Nuclear Information System (INIS)

1982-09-01

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

Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

International Nuclear Information System (INIS)

Estes, B.F.; Berry, D.T.

1980-02-01

Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

Gas cooled fast reactor background, facilities, industries and programmes

International Nuclear Information System (INIS)

Dalle Donne, M.

1980-05-01

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.) [de

Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1982-August 1983

International Nuclear Information System (INIS)

Brenizer, J.S.; Benneche, P.E.

1984-03-01

The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1982 was over 5500 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1982 through August 1983

Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1983-August 1984

International Nuclear Information System (INIS)

Mulder, R.U.; Benneche, P.E.

1984-11-01

The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1983 was over 6000 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1983 through August 1984

Research nuclear reactor RA - Annual Report 1990 with the comparative evaluation for the period 1986-1990

International Nuclear Information System (INIS)

Sotic, O.

1990-12-01

Annual report concerning the project 'RA research nuclear reactor' for 1990, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

2013-01-01

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

Accidents and failures in reactor facilities for test and research and reactor facilities in the stage of research and development in fiscal year 1987

International Nuclear Information System (INIS)

1988-01-01

The number of accidents and failures reported in fiscal year 1987 in conformity with the law on the regulation of nuclear reactors and others was three. One case occurred during operation, and two cases occurred in shutdown state. One case was caused by improper construction management, and two cases were due to improper maintenance management. The effect of radioactivity to the surrounding environment of reactor facilities due to these accidents and failures did not arise. These occurred in the NSRR of Japan Atomic Energy Research Institute (Tokai), the experimental FBR Joyo and the ATR Fugen Power Station of Power Reactor and Nuclear Fuel Development Corp. In addition to these, the light troubles reported on the basis of the notice from the director of Science and Technology Agency dated September 1, 1981, were three cases. (K.I.)

Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

International Nuclear Information System (INIS)

2011-01-01

Water is the most common fluid used to remove the heat produced in a research reactor (RR). It is also the most common media used to store spent fuel elements after being removed from the reactor core. Spent fuel is stored either in the at-reactor pool or in away-from-reactor wet facilities, where the fuel elements are maintained until submission to final disposal, or until the decay heat is low enough to allow migration to a dry storage facility. Maintaining high quality water is the most important factor in preventing degradation of aluminium clad fuel elements, and other structural components in water cooled research reactors. Excellent water quality in spent fuel wet storage facilities is essential to achieve optimum storage performance. Experience shows the remarkable success of many research reactors where the water chemistry has been well controlled. In these cases, aluminium clad fuel elements and aluminium pool liners show few, if any, signs of either localized or general corrosion, even after more than 30 years of exposure to research reactor water. In contrast, when water quality was allowed to degrade, the fuel clad and the structural parts of the reactor have been seriously corroded. The driving force to prepare this publication was the recognition that, even though a great deal of information on research reactor water quality is available in the open literature, no comprehensive report addressing the rationale of water quality management in research reactors has been published to date. This report is designed to provide a comprehensive catalogue of good practices for the management of water quality in research reactors. It also presents a brief description of the corrosion process that affects the components of a research reactor. Further, the report provides a basic understanding of water chemistry and its influence on the corrosion process; specifies requirements and operational limits for water purification systems of RRs; describes good practices

Operational experiences in radiation protection in fast reactor fuel reprocessing facility

International Nuclear Information System (INIS)

Meenakshisundaram, V.; Rajagopal, V.; Santhanam, R.; Baskar, S.; Madhusoodanan, U.; Chandrasekaran, S.; Balasundar, S.; Suresh, K.; Ajoy, K.C.; Dhanasekaran, A.; Akila, R.; Indira, R.

2008-01-01

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 85 Kr 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

A description of the demonstration Integral Fast Reactor fuel cycle facility

International Nuclear Information System (INIS)

Courtney, J.C.; Carnes, M.D.; Dwight, C.C.; Forrester, R.J.

1991-01-01

A fuel examination facility at the Idaho National Engineering Laboratory is being converted into a facility that will electrochemically process spent fuel. This is an important step in the demonstration of the Integral Fast Reactor concept being developed by Argonne National Laboratory. Renovations are designed to bring the facility up to current health and safety and environmental standards and to support its new mission. Improvements include the addition of high-reliability earthquake hardened off-gas and electrical power systems, the upgrading of radiological instrumentation, and the incorporation of advances in contamination control. A major task is the construction of a new equipment repair and decontamination facility in the basement of the building to support operations

1980 Annual status report reactor safety

International Nuclear Information System (INIS)

1981-01-01

The JRC reactor safety programme involves theoretical and experimental activities to analyse accidents and their consequences for LWRs and LMFBRs. The first project deals with the improvement and the application of methodologies for risk and reliability assessment. This activity involves the identification and modelling of accident sequences and events and the analysis of fault trees. In this project, the implementation of a centralized data bank system (European Reliability Data System) is foreseen, which should provide the information needed for risk assessment studies. In project 2 a major effort on LWRs is centered on the study of the loss-of-coolant accident following large, intermediate or small breaks of the primary circuit. These accidents are simulated out of pile in the LOBI facility. In project 3 a contribution is made to solve material problems and to provide data and calculation methods for end of life predictions of reactor components. It involves a contribution to the programme for the inspection of steel components (PISC) as well as the study of fracture and creep fatigue properties of stainless steel. In the project 4 and 5 a deterministic approach is adopted to solve some problems of large hypothetical accidents in an LMFBR. The calculation tools developed concern sodium thermohydraulics in fuel element bundles, fuel coolant interaction, whole core accident analysis, containment loading and response and post accident heat removal

2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

International Nuclear Information System (INIS)

Lewis, Mike

2012-01-01

This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

Reactor Engineering Division annual report

International Nuclear Information System (INIS)

1975-02-01

This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

Annual report ''nuclear safety in France''; Le rapport annuel ''la surete nucleaire en France''

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This document is the 2001 annual report of the French authority of nuclear safety (ASN). It summarizes the highlights of the year 2000 and details the following aspects: the nuclear safety in France, the organization of the control of nuclear safety, the regulation relative to basic nuclear facilities, the control of facilities, the information of the public, the international relations, the organisation of emergencies, the radiation protection, the transport of radioactive materials, the radioactive wastes, the PWR reactors, the experimental reactors and other laboratories and facilities, the nuclear fuel cycle facilities, and the shutdown and dismantling of nuclear facilities. (J.S.)

RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

1981-12-01

The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

Proposed power upgrade of the hot fuel examination facility's neutron radiography reactor

International Nuclear Information System (INIS)

Pruett, D.P.; Richards, W.J.; Heidel, C.C.

1984-01-01

The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both non-destructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the non-destructive examination techniques utilized at HFEF is neutron radiography. When the NRAD facility was designed and constructed, an operating power level of 250 kw was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. Since that time, several things have occurred that have tended to increase radiography exposure times to as much as 90 minutes each. In order to decrease exposure times, the reactor power level is to be increased from 250 kW to 1 MW. This increase in power will necessitate several engineering and design changes. The proposed upgrade of the NRAD facility will increase the neutron flux available in the beam tubes appreciably. The increased flux will enable NRAD to continue to meet its operational commitments in a timely manner and to develop state-of-the-art techniques in the future as it has in the past

In-pile experiments and test facilities proposed for fast reactor safety

International Nuclear Information System (INIS)

Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

1976-01-01

The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

International Nuclear Information System (INIS)

2000-01-01

Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

Preparation of an annual report for a consolidated radioisotope licence

International Nuclear Information System (INIS)

1985-01-01

A consolidated radioisotope licence is a single license issued by the Atomic Energy Control Board (AECB) to an institution having many users of radioactive materials. The licence is issued when the institution has fulfilled the requirements set by the AECB and has implemented policies and procedures which will ensure the maintenance of an effective radiation safety program. The consolidated licence is retained only if the results of AECB inspections are satisfactory -- or remedial action is taken promptly -- and if the institution reports regularly about the status of its radiation safety program. One aspect of this reporting procedure is an annual report. This guide describes the information that the AECB requires in the annual report. The guide applies primarily to universities and research institutions where a wide variety of radioisotope uses take place. The guide does not affect most other institutions or facilities, either because the nature and extent of their operations with radioactive materials do not lend themselves to consolidated licensing or because they are not licensed by the Radioisotopes and Transportation Division. (For example, this guide does not apply to reactors, to accelerators, or to waste management or uranium mining or refining facilities.) The information in the annual report should be specific to the consolidated radioisotope licence. Incidents, staff exposures, and waste associated with accelerators, research reactors, and waste management facilities should be excluded from this report unless they affected activities under the consolidated licence

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

International Nuclear Information System (INIS)

Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

2006-01-01

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

Radiation protection planning for decommissioning of research reactor facilities

International Nuclear Information System (INIS)

Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

2008-01-01

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)

RB research nuclear reactor, Annual report for 1981; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Sotic, O; Pesic, M; Vranic, S; Zivkovic, B; Bogdanovic, M; Petronijevic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1981-07-01

The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report.

Waste from decommissioning of research reactors and other small nuclear facilities

International Nuclear Information System (INIS)

Massaut, V.

2001-01-01

Full text: Small nuclear facilities were often built for research or pilot purposes. It includes the research reactors of various types and various aims (physics research, nuclear research, nuclear weapons development, materials testing reactor, isotope production, pilot plant, etc.) as well as laboratories, hot cells and accelerators used for a broad spectrum of research or production purposes. These installations are characterized not only by their size (reduced footprint) but also, and even mostly, by the very diversified type of materials, products and isotopes handled within these facilities. This large variety can sometimes enhance the difficulties encountered for the dismantling of such facilities. The presence of materials like beryllium, graphite, lead, PCBs, sodium, sometimes in relatively large quantities, are also challenges to be faced by the dismantlers of such facilities, because these types of waste are either toxic or no solutions are readily available for their conditioning or long term disposal. The paper will review what is currently done in different small nuclear facilities, and what are the remaining problems and challenges for future dismantling and waste management. The question of whether Research and Development for waste handling methods and processes is needed is still pending. Even for the dismantling operation itself, important improvements can be brought in the fields of characterization, decontamination, remote handling, etc. by further developments and innovative systems. The way of funding such facilities decommissioning will be reviewed as well as the very difficult cost estimation for such facilities, often one-of-a-kind. The aspects of radioprotection optimization (ALARA principle) and classical operators safety will also be highlighted, as well as the potential solutions or improvements. In fact, small nuclear facilities encounter often, when dismantling, the same problems as the large nuclear power plants, but have in

Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

International Nuclear Information System (INIS)

Calvaresi, C.; Moreschi, L.F.

1983-01-01

PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

Conceptual design of a fission-based integrated test facility for fusion reactor components

International Nuclear Information System (INIS)

Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

1982-01-01

The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

Annual Status Report (FY2016) Performance Assessment for the Environmental Restoration Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Casbon, M. A. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

2017-03-15

DOE O 435.1, Radioactive Waste Management, and DOE M 435.1-1, Radioactive Waste Management Manual, require that a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Annual summaries of low-level waste (LLW) disposal operations must be prepared with respect to the conclusions and recommendations of the PA and CA, and a determination of the need to revise the PA or CA must be made. The annual summary requirement provides a structured approach for demonstrating the continued adequacy of the PA and CA in demonstrating a reasonable expectation that the performance objectives will be met. This annual summary addresses only the status of the Environmental Restoration Disposal Facility (ERDF) PA (CP-60089, Performance Assessment for the Environmental Restoration Disposal Facility, Hanford Site, Washington, formerly WCH-520 Rev. 1)1. The CA for ERDF is supported by DOE/RL-2016-62, Annual Status Report (FY 2016): Composite Analysis of Low Level Waste Disposal in the Central Plateau at the Hanford Site. The ERDF PA portion of the CA document is found in Section 3.1.4, and the ERDF operations portion is found in Section 3.3.3.2 of that document.

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

International Nuclear Information System (INIS)

Moss, R.; Watkins, P.; Vroegindeweij, C.; Stecher-Rasmussen, F.; Huiskamp, R.; Ravensberg, K.; Appelman, K.; Sauerwein, W.; Hideghety, K.; Gabel, D.

2001-01-01

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)

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

International Nuclear Information System (INIS)

Jones, R.J.

1985-11-01

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

Personal neutron dosimetry at a research reactor facility

International Nuclear Information System (INIS)

Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

2001-01-01

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)

Facility for in-reactor creep testing of fuel cladding

International Nuclear Information System (INIS)

Kohn, E.; Wright, M.G.

1976-11-01

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)

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

International Nuclear Information System (INIS)

Trivedi, A.; Cornett, R.J.; Galeriu, D.; Workman, W.; Brown, R.M.

1997-02-01

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)

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

International Nuclear Information System (INIS)

Hofmeyr, C.; Mayer, R.M.; Tillwick, D.L.

1979-01-01

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-10UC 2 . (Auth.)

Concerning modification of plan for installation of reactors (addition of No.4 reactor and modification of No.1, No.2 and No.3 reactor facilities) at Hamaoka nuclear power plant of Chubu Electric Power Co., Ltd. (reply to inquiry)

International Nuclear Information System (INIS)

1988-01-01

The Nuclear Safety Commission on Oct. 22, 1987 directed the Nuclear Reactor Safety Expert Group to carry out a study on it, made deliberations after receiving a report from the Group on Jun. 28, and submitted the findings to the Minister of International Trade and Industry on Jul. 14. The study and deliberations were intended to determine the conformity of the modifications to the applicable laws relating to control of nuclear material, nuclear fuel and nuclear reactor. The investigation on the location covered the site conditions, geology, seismic environment, meteorology, hydrology, and social environment. The investigations on the safety design of the reactor facilities addressed the design of the overall reactor facilities, anti-earthquake design, and design of other equipment including reactor core, fuel equipment, monitors, reactivity controllers, safety protection equipment, and emergency core cooler. Other investigations included exposure evaluation and accident analysis. It was concluded that the modifications would not have adverse effects on the safety of the reactor facilities. (Nogami, K.)

Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

Energy Technology Data Exchange (ETDEWEB)

D. E. Lewis D. A. Hagemeyer Y. U. McCormick

2012-07-07

This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor

Atomic Energy of Canada Limited annual report 1999-2000

International Nuclear Information System (INIS)

2000-01-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2000, and summarizes the activities of AECL during the period 1999-2000. The activities covered in this report include the CANDU reactor business, with the completion of the Wolsong unit 4 in the Republic of Korea, progress in the construction of two CANDU reactors for the Qinshan CANDU project in China, as well as the service business with Ontario Power Generation in the rehabilitation and life extension of operating CANDU reactors. In the R and D programs there is on-going effort towards the next generation of reactor technologies for CANDU nuclear power plants, discussions continue on the funding for the Canadian Neutron Facility for materials research (CNF) and progress being made on the Maple medical isotope reactor

Atomic Energy of Canada Limited annual report 1999-2000

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2000, and summarizes the activities of AECL during the period 1999-2000. The activities covered in this report include the CANDU reactor business, with the completion of the Wolsong unit 4 in the Republic of Korea, progress in the construction of two CANDU reactors for the Qinshan CANDU project in China, as well as the service business with Ontario Power Generation in the rehabilitation and life extension of operating CANDU reactors. In the R and D programs there is on-going effort towards the next generation of reactor technologies for CANDU nuclear power plants, discussions continue on the funding for the Canadian Neutron Facility for materials research (CNF) and progress being made on the Maple medical isotope reactor.

Technical meeting (TM) to 'Review of national programmes on fast reactors and accelerator driven systems (ADS)'. Technical Working Group on Fast Reactors (TWG-FR) (37th annual meeting). Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The objectives of the 37th Annual Meeting of the Technical Working Group on Fast Reactors, were to: 1) exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); 2) review the progress since the 36th TWG-FR Annual Meeting, including the status of the actions; 3) consider meeting arrangements for 2004 and 2005; 4) review the Agency's co-ordinated research activities in the field of FRs and ADS, as well as co-ordination of the TWG-FR's activities with other organizations. The participants made presentations on the status of the respective national programmes on FR and ADS development. A summary of the highlights for the period since the 36th TWG-FR Annual Meeting is included in this proceedings. Annex IV contains the Review of National Programs on Fast Reactors and Accelerator Driven Systems (ADS), and the TWG-FR Activity Report for the Period May 2003-April 2004.

Refueling the RPI reactor facility with low-enrichment fuel

International Nuclear Information System (INIS)

Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

1985-01-01

The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

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

International Nuclear Information System (INIS)

Zaffuts, P. J.

2003-01-01

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

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.

Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

International Nuclear Information System (INIS)

Rana, Y.S.; Mishra, Abhishek; Singh, Tej

2016-06-01

Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

International Nuclear Information System (INIS)

Seidensticker, R.W.

1989-01-01

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. This paper reviews the research and development (R and 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 and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

Status of national programmes on fast breeder reactors. Eighteenth annual meeting, Vienna, Austria, 16-19 April 1985

International Nuclear Information System (INIS)

1986-02-01

The Eighteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in April 1985. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A separate abstract was prepared for each of the 12 presentations of the meeting

GKSS. Annual report 2005/2006. Science use

International Nuclear Information System (INIS)

2006-01-01

The annual report of the GKSS research centers presents selected activities from the research program: light weight structures for transportation and energy, synchrotron radiation for materials research development of x-ray mirrors for VUV-free electron lasers, membranes in process- and biomedical technology, neutralization of acid lakes, membranes for bioartificial organs, the coastal environment, maxwave project, mercury deposition fluxes into arctic and antarctic regions, monitoring systems for coastal regions, research reactor, training facilities. (GL)

2016 Annual Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Ponds

International Nuclear Information System (INIS)

Lewis, Michael George

2017-01-01

This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2015-October 31, 2016. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019. This report contains the following information: · Facility and system description · Permit required effluent monitoring data and loading rates · Permit required groundwater monitoring data · Status of compliance activities · Issues · Discussion of the facility's environmental impacts. During the 2016 permit year, 180.99 million gallons of wastewater were discharged to the Cold Waste Ponds. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest in well USGS-065, which is the closest downgradient well to the Cold Waste Ponds. Sulfate and total dissolved solids concentrations decrease rapidly as the distance downgradient from the Cold Waste Ponds increases. Although concentrations of sulfate and total dissolved solids are significantly higher in well USGS-065 than in the other monitoring wells, both parameters remained below the Ground Water Quality Rule Secondary Constituent Standards in well USGS-065. The facility was in compliance with the Reuse Permit during the 2016 permit year.

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

International Nuclear Information System (INIS)

Israr, M.; Shami, Qamar-ud-din; Pervez, S.

1997-11-01

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)

Gamma exposure rate estimation in irradiation facilities of nuclear research reactors

International Nuclear Information System (INIS)

Daoud, Adrian

2009-01-01

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)

1997 Annual report

International Nuclear Information System (INIS)

1997-01-01

This annual report present information of the main activities on the scope of the radiation protection and nuclear safety of the Nuclear Regulatory Authority (ARN) of the Argentina during 1997. The activities are developed in 11 chapters and 1 appendix, were are descripting the activities and the bases of the Argentine Regulatory System. The following chapters and the appendix are: 1.-Evolution of the nuclear regulatory activities in the Argentina. 2.-Regulatory system, laws and regulations. 3.-Institutional relations with national and international organizations. 4.-Inspections and evaluations of the nuclear reactors. 5.-Inspections and evaluations of the radioactive facilities. 6.-Inspections and evaluations of safeguards and physical protection. 7.-Radiation control of the occupational and public doses. 8.-Radiological emergencies. 9.-Projects and tasks scientific-technological. 10.-Training courses and meetings. 11.-Economic and human recourses. Appendix.- Inspections of radioactive facilities

Fission reactor based epithermal neutron irradiation facilities for routine clinical application in BNCT-Hatanaka memorial lecture

International Nuclear Information System (INIS)

Harling, Otto K.

2009-01-01

Based on experience gained in the recent clinical studies at MIT/Harvard, the desirable characteristics of epithermal neutron irradiation facilities for eventual routine clinical BNCT are suggested. A discussion of two approaches to using fission reactors for epithermal neutron BNCT is provided. This is followed by specific suggestions for the performance and features needed for high throughput clinical BNCT. An example of a current state-of-the-art, reactor based facility, suited for routine clinical use is discussed. Some comments are provided on the current status of reactor versus accelerator based epithermal neutron sources for BNCT. This paper concludes with a summary and a few personal observations on BNCT by the author.

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

Science.gov (United States)

2012-05-03

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

Atomic Energy of Canada Limited annual report 2000-2001

International Nuclear Information System (INIS)

2001-01-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor

Atomic Energy of Canada Limited annual report 2000-2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This is the annual report of the Atomic Energy of Canada Limited for the year ending March 31, 2001 and summarizes the activities of AECL during the period 2000-2001. The activities covered in this report include the CANDU reactor business, with progress being reported in the construction of two CANDU 6 reactors for the Qinshan CANDU project in China, the anticipated completion of Cernavoda unit 2, the completion of spent fuel storage at Cernavoda unit 1 in Romania, as well as the service business with New Brunswick Power, Ontario Power Generation, Bruce Power and Hydro Quebec in the refurbishment of operating, CANDU reactors. In the R and D programs discussions continue on funding for the Canadian Neutron Facility for Materials Research (CNF) and progress on the Maple medical isotope reactor.

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

International Nuclear Information System (INIS)

Salazar, M.D.

1998-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Salazar, M.D.

1998-12-01

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.

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Present state of inspection robot technology in nuclear power facilities. Case of fast breeder reactors

International Nuclear Information System (INIS)

Ara, Kuniaki

1995-01-01

In the maintenance works in nuclear power facilities such as checkup, inspection and repair, for the main purpose of radiation protection, remote operation technology was introduced since relatively early stage, and at present, the robots that carry out the inspection works for confirming the soundness of main equipment have been developed and put to practical use. At the time of introducing these technologies, in addition to the research and development of robots proper, the coordination with the design of plant machinery and equipment facilities as the premise of introducing robots is an important requirement. In this report, the present state of the development of remote inspection technology for fast breeder reactors is introduced, and the matters to which attention is paid in the plant design for introducing robots are explained. First, fast breeder reactors are described. The needs of robotizing and adopting remote operation in nuclear power facilities are explained, using the examples of the inspection system for a reactor vessel and the inspection system for steam generator heat transfer tubes. (K.I.)

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

International Nuclear Information System (INIS)

Malkoske, G.R.

2002-01-01

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

Research nuclear reactor RA - Annual Report 1995, with comparative review for period 1991-1995

International Nuclear Information System (INIS)

Sotic, O.

1995-12-01

Activities related to revitalisation of the RA reactor started in 1986, were continued during period 1991-1995. All the planned actions related to renewal of the reactor components were finished. The last, and at the same time most important action, related to exchange of complete reactor instrumentation is underway, delayed. The delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 47 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved during the period 1991-1995. This research reactor RA Annual is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

Annual report of the research works with joint-use JAERI facilities for fiscal 1974

International Nuclear Information System (INIS)

1975-01-01

Results of the research works by national universities with JAERI's (Japan Atomic Energy Research Institute) joint-use facilities for fiscal 1974 are described. Facilities are research reactors, Co-60, Linac, etc. Research results are presented in individual summaries, covering radiation damage and solid state physics, activation analysis and nuclear chemistry, irradiation effects, etc. Results of the joint works with JAERI are also presented similarly. (Mori, K.)

Annual report of the Neutron Irradiation and Testing Reactor Center. FY2007. April 1, 2007 - March 31, 2008

International Nuclear Information System (INIS)

2009-03-01

The Japan Materials Testing Reactor (JMTR), achieving first criticality in March 1968, has been used in testing the durability and integrity of reactor fuels and components, basic nuclear research, the production of radioisotopes (RIs), and other purposes. The JMTR, however, stopped in August 2006 after its 165th operation cycle, and is currently under going partial renewal of reactor facilities and installation of new irradiation Facilities, geared toward being restarted in 2011. In addition, to cope with the strong requests from users to improve usability of the JMTR, efforts are being made to increase reactor operation efficiency, shorten the turnaround time for obtaining results, and other necessary tasks for JMTR to commence reoperation. The present report summarizes the activities carried out in 2007 for the refurbishment and restart of JMTR. (author)

Handling of multiassembly sealed baskets between reactor storage and a remote handling facility

International Nuclear Information System (INIS)

Massey, J.V.; Kessler, J.H.; McSherry, A.J.

1989-06-01

The storage of multiple fuel assemblies in sealed (welded) dry storage baskets is gaining increasing use to augment at-reactor fuel storage capacity. Since this increasing use will place a significant number of such baskets on reactor sites, some initial downstream planning for their future handling scenarios for retrieving multi-assembly sealed baskets (MSBs) from onsite storage and transferring and shipping the fuel (and/or the baskets) to a federally operated remote handling facility (RHF). Numerous options or at-reactor and away-from-reactor handling were investigated. Materials handling flowsheets were developed along with conceptual designs for the equipment and tools required to handle and open the MSBs. The handling options were evaluated and compared to a reference case, fuel handling sequence (i.e., fuel assemblies are taken from the fuel pool, shipped to a receiving and handling facility and placed into interim storage). The main parameters analyzed are throughout, radiation dose burden and cost. In addition to evaluating the handling of MSBs, this work also evaluated handling consolidated fuel canisters (CFCs). In summary, the handling of MSBs and CFCs in the store, ship and bury fuel cycle was found to be feasible and, under some conditions, to offer significant benefits in terms of throughput, cost and safety. 14 refs., 20 figs., 24 tabs

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

International Nuclear Information System (INIS)

Bailey, E.N.

2008-01-01

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

Annual monitoring and surveillance report for Piqua Nuclear Power Facility, Piqua, Ohio

International Nuclear Information System (INIS)

Mosho, G.D.

1991-12-01

This report discusses the decommissioned Piqua Nuclear Power Facility which is located in Piqua, Ohio near the Greater Miami River. The Facility was built by the US Atomic Energy Commission (now U. S. Department of Energy) and was operated from 1963 to 1966. The reactor was retired prior to 1970 and the facility was leased to the city of Piqua for use as offices and equipment storage. In December 1991, a radiological survey was done of the facility to document its radiological condition. The data show that all radiological parameters measured were essentially the same as that found in the natural environment. The only exception was that low levels of radioactive contamination were detected in one drain on the 56.5 ft elevation, but the radiation exposure rate in that area was also typical of natural background

RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

Energy Technology Data Exchange (ETDEWEB)

Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1984-07-01

The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

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

International Nuclear Information System (INIS)

Miller, C.W.; Meyer, H.R.

1985-01-01

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

The upgrading of the cyclic neutron activation analysis facility at the Dalat research reactor

International Nuclear Information System (INIS)

Van Doanh Ho; Manh Dung Ho; Quang Thien Tran; Dong Vu Cao; Thanh Viet Ha

2018-01-01

The cyclic neutron activation analysis (CNAA) facility based on a pneumatic transfer system for short irradiation and rapid counting has recently been upgraded at the Dalat research reactor. The original facility was only designed for single irradiation. Therefore, this work has aimed to upgrade both hardware and software for the cyclic irradiation. In this paper, the upgrading of the facility for CNAA was described. Irradiation time of the facility were calibrated, thereby reducing irradiation time to seconds with precision. The accuracy and sensitivity of CNAA based-on the upgraded facility were assessed by determination of some short-lived nuclides. (author)

Operation and maintenance of the RB reactor, Annual report for 1977; Pogon i odrzavanje reaktora RB, Izvestaj o radu u 1977. godini

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1977-07-01

The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators.