The application of modern safety criteria to restarting and operating the USDOE K-Reactor

International Nuclear Information System (INIS)

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-01-01

The United States Department of Energy's (USDOE's) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992

Advanced Test Reactor outage risk assessment

International Nuclear Information System (INIS)

Thatcher, T.A.; Atkinson, S.A.

1997-01-01

Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-05-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1983-01-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1983-03-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-11-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-10-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-08-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-09-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Study on Operator Actions during the Occurrences of Undesirable Events in PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Tom, P.P.; Nurul Husna Zainal Abidin; Lanyau, T.A.; Zaredah Hashim

2016-01-01

Due to the recent Fukushima accident, the potential risks at one and only nuclear research reactor in the country, which is the PUSPATI TRIGA Reactor (RTP), has increasingly gain concerns and an attempt on the development of Level 1 Probabilistic Safety Assessment (PSA) for this reactor has been commenced. The preliminary scope of the PSA is to analyse the risk of core degradation during normal daily operation due to the random component failure and human error. SPAR-H and THERP method is used for quantifying human error probability (HEP). However, the scopes of this study only cover the qualitative parts that use interview/questionnaire method. The objectives of the questionnaire are to identify the main action for RTP operators when any undesired incident occurs during full power operation that might be caused by random component failures. From the questionnaires that have been conducted, the respondents consisted of 4 licensed operators and 9 trainee operators. All licensed operators have experience of operating reactor for more than 15 years while the trainee operator have been operate the reactor with experience of less than 10 years. Generally, in the event of an abnormal condition involving the reactor, an operator whether a licensed operator or the trainee does not have to ask permission in advance from the top individuals to carry out scram. This is to prevent the situation becoming increasingly severe if the reactor is still operating. With complete training and knowledge derived from the management, an operator can act efficiently in any emergency case. (author)

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

Reactor operational transient analysis

International Nuclear Information System (INIS)

Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

1983-01-01

To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

Linguistic Formalism for Semi-Autonomous Reactor Operation

International Nuclear Information System (INIS)

Joo, Sungmoon; Seo, Sang Mun; Suh, Yong-Suk; Park, Cheol

2017-01-01

The ultimate goal of our work is to develop a novel, integrated system for semi-autonomous reactor operation by introducing an interfacing language shared by human reactor operators and artificially intelligent service agents (e.g., robots). We envision that human operators and artificially intelligent service agents operate the reactor cooperatively in the future. For example, an artificially intelligent service agent carries out a human reactor operator's command or reports the result of a task commanded by the human reactor operator. This work presents preliminary work towards a unified linguistic formalism for cooperative, semiautonomous reactor operation. Application of the proposed formalism to reactor operator communication domain shows that the formalism effectively captures the syntax and semantics of the domain-specific language defined by the communication protocol.

Reactor core operation management system

International Nuclear Information System (INIS)

Sato, Tomomi.

1992-01-01

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

Reactor core operation management system

Energy Technology Data Exchange (ETDEWEB)

Sato, Tomomi.

1992-05-28

Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1981-08-01

The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

Operating reactors licensing actions summary

International Nuclear Information System (INIS)

1982-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

Reactor operation safety information document

Energy Technology Data Exchange (ETDEWEB)

1990-01-01

The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

Operating US power reactors

International Nuclear Information System (INIS)

Silver, E.G.

1988-01-01

This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

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

Reactor operations at SAFARI-1

International Nuclear Information System (INIS)

Vlok, J.W.H.

2003-01-01

A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Report of the reactor Operators Service - Annex F

International Nuclear Information System (INIS)

Zivotic, Z.

1992-01-01

RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. For continuous training of the existing operator staff the Service has prepared and published eleven booklets: Nuclear reactor; RA reactor primary coolant loop; System for purification of heavy water; reactor helium system; system for technical water; electric power system; control and operation; ventilation system in the reactor building; special sewage system; construction properties of the reactor core; reactor building and installations. During the reporting period there have been no accidents nor incidents that could affect the reactor personnel [sr

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

1989-12-01

The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Management of safety and risk at the HFIR [High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Glovier, H.A.

1990-01-01

This paper discusses the management of safety and risk at the High-Flux Isotope Reactor (HFIR), a category A research reactor at Oak Ridge National Laboratory (ORNL). The HFIR went critical in 1966 and operated at its designed 100 MW for 20 yr until it was shut down on November 14, 1986. It operated at a >90% availability and without significant event during this period. The result was a complacent management program lacking rigor. This complacency came to an end with the Chernobyl accident, which led to the appointment of an internal committee to assess the safety of ORNL reactor operations. This committee found that HFIR pressure vessel material specimens removed several years earlier had not been analyzed. This issue led to a general review of management practices that were found lacking in quality assurance, safety documentation, training process, and emergency planning, among others. Management accountability was lacking, as shown by design basis and safety analyses that were not up to data and by the fact that reactor operators whose requalification examinations had not been graded were allowed to continue operating the reactor over a long period of time. Between shutdown in 1986 and restart in April 1989, significant management changes and initiatives were made in the area of risk and safety management of ORNL reactors. These are presented briefly in this paper

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Siting analysis and risk assessment for small single-purpose heating reactors

International Nuclear Information System (INIS)

Tarjanne, R.

1979-04-01

Two alternative sites both 10km away from the centre of Helsinki are considered for reactor unit sizes of 400mw and 800mw. The risks associated with a small single-purpose heating reactor is evaluated for normal operation and accident conditions. The evaluation for accident condition is performed for three characteristics accidents. Three pathways are considered in the calculation of the radiation exposure: direct external gamma dose from the release plume, direct gamma radiation from deposited activity on the ground and dose due to inhalation. The risks are compared with the risks from alternative conventional fossil fuelled district heat production methods. The results show that the heating reactor alternative causes an unsignificant risk, which is far less than the risk caused by the fossil-fuelled alternatives

Method of operating a nuclear reactor

International Nuclear Information System (INIS)

Spurgin, A.J.; Schaefer, W.F.

1978-01-01

A method of controlling a nuclear power generting station in the event of a malfunction of particular operating components is described. Upon identification of a malfunction, preselected groups of control rods are fully inserted sequentially until a predetermined power level is approached. Additional control rods are then selectively inserted to quickly bring the reactor to a second given power level to be compatible with safe operation of the system with the malfunctioning component. At the time the thermal power output of the reactor is being reduced, the turbine is operated at a rate consistent with the output of the reactor. In the event of a malfunction, the power generating system is operated in a turbine following reactor mode, with the reactor power rapidly reduced, in a controlled manner, to a safe level compatible with the type of malfunction experienced

Risk-informed design of a pebble bed gas reactor

International Nuclear Information System (INIS)

Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

2003-01-01

One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

Impact of proposed research reactor standards on reactor operation

Energy Technology Data Exchange (ETDEWEB)

Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

1974-07-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Impact of proposed research reactor standards on reactor operation

International Nuclear Information System (INIS)

Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

1974-01-01

A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

Method of safely operating nuclear reactor

International Nuclear Information System (INIS)

Ochiai, Kanehiro.

1976-01-01

Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)

Extended life for the Borssele reactor. Risks and cost

International Nuclear Information System (INIS)

Van Gelder, J.W.; De Rijk, P.

2005-01-01

In the last few years several decisions were taken by the Dutch government with respect to the service life of the nuclear power reactor Borssele. The last decision was taken in 2003 to close the reactor in the year 2013. However, the discussion started again while it appears, according to the Dutch Minister of State for the Environment, there is not enough political and legal basis to enforce closure of the reactor in 2013. In this report insight is given into the risks and societal cost to continue the operation of the Borssele reactor after 2013 [nl

Health requirements for nuclear reactor operators

International Nuclear Information System (INIS)

1980-05-01

The health prerequisites established for the qualification of nuclear reactor operators according to CNEN-NE-1.01 Guidelines Licensing of nuclear reactor operators, CNEN-12/79 Resolution, are described. (M.A.) [pt

Review and updates of the risk assessment for advanced test reactor operations for operating events and experience

International Nuclear Information System (INIS)

Atkinson, S.A.

1996-01-01

Annual or biannual reviews of the operating history of the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) have been conducted for the purpose of reviewing and updating the ATR probabilistic safety assessment (PSA) for operating events and operating experience since the first compilation of plant- specific experience data for the ATR PSA which included data for operation from initial power operation in 1969 through 1988. This technical paper briefly discusses the means and some results of these periodic reviews of operating experience and their influence on the ATR PSA

Digital computer operation of a nuclear reactor

International Nuclear Information System (INIS)

Colley, R.W.

1984-01-01

A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

N reactor individual risk comparison to quantitative nuclear safety goals

International Nuclear Information System (INIS)

Wang, O.S.; Rainey, T.E.; Zentner, M.D.

1990-01-01

A full-scope level III probabilistic risk assessment (PRA) has been completed for N reactor, a US Department of Energy (DOE) production reactor located on the Hanford Reservation in the state of Washington. Sandia National Laboratories (SNL) provided the technical leadership for this work, using the state-of-the-art NUREG-1150 methodology developed for the US Nuclear Regulatory Commission (NRC). The main objectives of this effort were to assess the risks to the public and to the on-site workers posed by the operation of N reactor, to identify changes to the plant that could reduce the overall risk, and to compare those risks to the proposed NRC and DOE quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford Company (WHC) and SNL for individual health risk evaluation, its results, and a comparison to the NRC safety objectives and the DOE nuclear safety guidelines. The N reactor results, are also compared with the five NUREG-1150 nuclear plants. Only internal events are compared here because external events are not yet reported in the current draft NUREG-1150. This is the first full-scope level III PRA study with a detailed quantitative safety goal comparison performed for DOE production reactors

Operational reactor physics analysis codes (ORPAC)

International Nuclear Information System (INIS)

Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

2007-07-01

For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

Method of operating a reactor

International Nuclear Information System (INIS)

Oosumi, Katsumi; Yamamoto, Michiyoshi.

1980-01-01

Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

Risk-based management system development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Davis, M.L.; Eide, S.A.

1990-01-01

A Risk-Based Management System (RBMS) is being developed to facilitate the use of the Advanced Test Reactor (ATR) probabilistic risk assessment to support ATR operation. Most ATR RBMS questions can best be answered using the System Analysis and Risk Assessment System (SARA) developed at the Idaho National Engineering Laboratory. However, some applications may require employment of the other four codes used to develop and report the PRA. These four codes include the Integrated Reliability and Risk Analysis System (IRRAS), SETS, ETA-II, and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The ATR RBMS will evolve over three years, and will include the results of the Level 3 and external events analysis

Risk assessment for Argon-41 in RP-10 research reactor

International Nuclear Information System (INIS)

Arenas C, J.; Alarcon V, P.

1987-01-01

An evaluation of the radiological risk during normal operations of the Peruvian 10 Mw research reactor was performed. The noble gas, Argon-41 produced from activation of Ar-40 from the air contained in the irradiation tubes, thermal column and coolant, is analyzed as the major source of risk

Reactor operation monitor

International Nuclear Information System (INIS)

Sakagami, Masaharu.

1982-01-01

Purpose: To improve the working performance of a reactor by extending the range for the power conditioning due to the control rod operation and flow rate control. Constitution: The results of calculations for the power distribution and the burn-up degree distribution of the reactor core from a reactor performance computer that processes each of measuring signals in a nuclear power plant are used as the inputs for a computing device of the fuel rod power hysteresis to form the power hysteresis for each of the fuel rods up to the present time. The data are used as the inputs for the computing device of the fuel rod performance index, and the fuel rod performance index representing the critical values for the stresses in the fuel rod cladding tubes and the critical values for the duration of the stresses determined from the power hysteresis and the burn-up degree of the fuel rod are calculated for each of the fuel rods. Accordingly, the power conditioning can be carried out upon power-up in the reactor while monitoring the fuel rod performance index f(t) for each of the fuel assemblies, whereby the range for the power conditioning due to the control rod operation and the flow rate control can be extended relative to fuel assemblies in which f(t) is smaller than 1. (Yoshino, Y.)

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

Reactor operation feed-back in France

International Nuclear Information System (INIS)

Feltin, C.; Fourest, B.; Libmann, J.

1982-09-01

The Nuclear Safety Department (DSN), technical support of French Safety Authorities, is, in particular, in charge of the analysis of reactor operation and of measures taken consequently to incidents. It proposed the criteria used to select significant incidents; it analyzes such incidents. DSN also analyzes the operating experience of each plant, several years after starting. It examines foreign incidents to assess in what extent lessons learned can be applied to french reactors. The examples presented show that to improve the safety of units operation, the experience feed-back leads to make arrangements, or modifications concerning not only circuits or materials but often procedures. Moreover they show the importance of procedures concerning the operations carried out during reactor shutdown

Method of reactor operation

International Nuclear Information System (INIS)

Maeda, Katsuji.

1982-01-01

Purpose: To prevent stress corrosion cracks in stainless steels caused from hydrogen peroxide in reactor operation in which the density of hydrogen peroxide in the reactor water is controlled upon reactor start-up. Method: A heat exchanger equipped with a heat source for applying external heat is disposed into the recycling system for reactor coolants. Upon reactor start-up, the coolants are heated by the heat exchanger till arriving at a temperature at which the dissolving rate is faster than the forming rate of hydrogen peroxide in the coolants, and nuclear heating is started after reaching the above temperature. The temperature of the reactor water is increased in such a manner and, when it arrives at 140 0 C, extraction of control elements is started and the heat source for the heat exchanger is interrupted simultaneously. In this way spikes in the density of hydrogen peroxide are suppressed upon reactor start-up to thereby decrease the stress corrosion cracks in stainless steels. (Horiuchi, T.)

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

Regulation for installation and operation of marine reactors

International Nuclear Information System (INIS)

1979-01-01

The regulation is defined under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and the provisions of the order for execution of the law. The regulation is applied to marine reactors and reactors installed in foreign nuclear ships. Basic concepts and terms are explained, such as: radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area and employee. The application for permission of installation of reactors shall list maximum continuous thermal power, location and general structure of reactor facilities, structure and equipment of reactors and treatment and storage facilities of nuclear fuel materials, etc. The application for permission of reactors installed in foreign ships shall describe specified matters according to the provisions for domestic reactors. The operation program of reactors for three years shall be filed to the Minister of Transportation for each reactor every fiscal year from that year when the operation is expected to start. Records shall be made for each reactor and kept for particular periods on inspection of reactor facilities, operation, fuel assembly, control of radiation, maintenance and others. Exposure doses, inspection and check up of reactor facilities, operation of reactors, transport and storage of nuclear fuel materials, etc. are designated in detail. (Okada, K.)

Computer monitoring of the RB reactor operation

International Nuclear Information System (INIS)

Milovanovic, S.; Pesic, M.; Milovanovic, T.

1998-01-01

Personal computer based acquisition system designed for monitoring of operation of the RB experimental reactor in the Institute of Nuclear Sciences 'Vinca' (former 'Boris Kidric') and experiences acquired during its use are shown in this paper. The monitoring covers generally all nuclear aspects of the reactor operation (start-up, nominal power operation, power changing, shut down and maintenance), but the emphasis is put on: real time (especially fast changing) reactivity measurement; supervising time dependence of the safety rods positions during shut down, and detection of position inaccuracy or failure operation of safety/control rods during the reactor operation or maintenance. (author)

Method of operating heavy water moderated reactors

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1980-01-01

Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

Licensed operating reactors

International Nuclear Information System (INIS)

1989-11-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page

Research about reactor operator's personality characteristics and performance

International Nuclear Information System (INIS)

Wei Li; He Xuhong; Zhao Bingquan

2003-01-01

To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

Preparation fo nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

Preparation of nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs

Standards for safe operation of research reactors

International Nuclear Information System (INIS)

1996-01-01

The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

Development of a risk monitor for research reactor

International Nuclear Information System (INIS)

Agrawal, Mayank; Mathur, D.; Joshi, N.S.; Varde, P.V.

2014-01-01

The traditional approach to engineering asset management, which primarily includes management and operation and maintenance activities, in a nuclear plant is based on engineering judgment. Here, the defence-in-depth and conservative criteria form the basis of decisions. Ideally, it is expected that the decisions should be based on performance of systems, structures and components (SSC), which should also take plant safety margin into consideration. Probabilistic methods are widely used as a compliment to deterministic methods in assessing safety of nuclear reactors. A risk-based operations and maintenance management system has been developed for a research reactor. The results of a level-1 PSA study in terms of minimal cut-set equations along with other engineering plant information forms major input for this system. The system is expected to be operational in plant control room conditions in support of routine decisions. This paper presents one of the main applications of the plant Level-1 Probabilistic Safety Assessment, developed from plant specific data i.e. a risk monitor. The risk monitor has several applications that support in decision making. It is expected that the use of this system in real-time scenario will provide improved understanding of safety on one hand while enabling enhancement of plant economics on the other. Some the case studies are discussed here. (author)

Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

International Nuclear Information System (INIS)

Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

2000-01-01

Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

Twenty years of health physics research reactor operation

International Nuclear Information System (INIS)

Sims, C.S.; Gilley, L.W.

1983-01-01

The Health Physics Research Reactor at the Oak Ridge National Laboratory has been in regular use for more than two decades. Safe operation of this fast reactor over this extended period indicates that (1) fundamental design, (2) operational procedures, (3) operator training and performance, (4) maintenance activites, and (5) management have all been eminently satisfactory. The reactor and its uses are described, the operational history and significant events are reviewed, and operational improvements and maintenance are discussed

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

Operation management of the prototype heavy water reactor 'Fugen'

International Nuclear Information System (INIS)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

1983-01-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

Operation and maintenance of 1MW PUSPATI TRIGA reactor

International Nuclear Information System (INIS)

Adnan Bokhari; Mohammad Suhaimi Kassim

2006-01-01

The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

The qualification of reactor operators

International Nuclear Information System (INIS)

Lima, J.M. de; Soares, H.V.

1981-01-01

The qualification and performance of nuclear power personnel have an important influence on the availability and safety operation of these plants. This paper describes the Brazilian rules and norms established by the CNEN-Brazilian Atomic Energy Comission, as well as policy of other countries concerning training requirements and experiences of nuclear power reactor operators. Some coments are made about the im pact of the march 1979 Three Mile Island accident on upgrading the reactor training requirements in U.S.A. and its international implication. (Author) [pt

Neutronics calculations for denatured molten salt reactors: Assessing resource requirements and proliferation-risk attributes

International Nuclear Information System (INIS)

Ahmad, Ali; McClamrock, Edward B.; Glaser, Alexander

2015-01-01

Highlights: • We study the proliferation-risk and resource attributes of denatured MSRs. • MSRs offer significantly better resource efficiency compared to light-water reactors. • Denatured single-fluid MSRs reactors offer promising non-proliferation attributes. - Abstract: Molten salt reactors (MSRs) are often advocated as a radical but worthwhile alternative to traditional reactor concepts based on solid fuels. This article builds upon the existing research into MSRs to model and simulate the operation of thorium-fueled single-fluid and two-fluid reactors. The analysis is based on neutronics calculations and focuses on denatured MSR systems. Resource utilization and basic proliferation-risk attributes are compared to those of standard light-water reactors. Depending on specific design choices, even fully denatured reactors could reduce uranium and enrichment requirements by a factor of 3–4. Overall, denatured single-fluid designs appear as the most promising candidate technology minimizing both design complexity and overall proliferation risks despite being somewhat less attractive from the perspective of resource utilization

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Regulation for installation and operation of experimental-research reactor

International Nuclear Information System (INIS)

1979-01-01

The ordinance is stipulated under the Law for regulation of nuclear raw materials, nuclear fuel materials and reactors and the provisions for installation and operation of reactor in the order for execution of the law. Basic concepts and terms are defined, such as, radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; preserved area; inspected surrounding area and employee. An application for permission of installation of reactor shall list such matters as: the maximum continuous thermal output of reactor; location and general construction of reactor facilities; construction and equipment of the main reactor and other facilities for nuclear fuel materials; cooling and controlling system and radioactive waste, etc. An operation plan of reactor for three years shall be filed till January 31 of the fiscal year preceding that one the operation begins. Records shall be made and kept for specified periods respectively on inspection of reactor facilities, operation, fuel assembly, radiation control, maintenance, accidents of reactor equipment and weather. Detailed rules are settled for entrance limitation to controlled area, exposure dose, inspection, check up and regular independent examination of reactor facilities, operation of reactor, transportation of substances contaminated by nuclear fuel materials within the works and storage, etc. (Okada, K.)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Test reactor risk assessment methodology

International Nuclear Information System (INIS)

Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

1976-04-01

A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

Development of a risk-based inservice inspection program for a liquid metal reactor

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.

1996-01-01

The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

Energy Technology Data Exchange (ETDEWEB)

Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

2013-04-04

Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

PUSPATI Triga Reactor - First year in operation

International Nuclear Information System (INIS)

Nahrul Khair Rashid.

1983-01-01

First year operation of RTP reactor was mostly devoted to making in house training, setting up and testing the facilities in preparation for more routine operations. Generally the operations are categorized into 4 main purposes; experiment of research, teaching and training, demonstration, and testing and maintenance. These four purposes are elaborated in detail. Additions and modifications were performed in order to improve the safety of reactor operation. (A.J.)

Reactor Operations informal monthly report December 1994

International Nuclear Information System (INIS)

1994-12-01

Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included

Risk analysis of releases from accidents during mid-loop operation at Surry

International Nuclear Information System (INIS)

Jo, J.; Lin, C.C.; Nimnual, S.; Mubayi, V.; Neymotin, L.

1992-11-01

Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these conditions: One at the Brookhaven National Laboratory for the Surry plant, a pressurized water reactor (PWR), and the other at the Sandia National Laboratories for the Grand Gulf plant, a boiling water reactor (BWR). Each of the studies consists of three linked, but distinct, components: a Level I probabilistic risk analysis (PRA) of the initiating events, systems analysis, and accident sequences leading to core damage; a Level 2/3 analysis of accident progression, fuel damage, releases, containment performance, source term and consequences-off-site and on-site; and a detailed Human Reliability Analysis (HRA) of actions relevant to plant conditions during LP/S operations. This paper summarizes the approach taken for the Level 2/3 analysis at Surry and provides preliminary results on the risk of releases and consequences for one plant operating state, mid-loop operation, during shutdown

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

Risk-informing special treatment requirements for reactors

International Nuclear Information System (INIS)

McKenna, E.M.; Reed, T.A.

2001-01-01

The U.S. Nuclear Regulatory Commission (NRC) is proposing to make regulatory changes to the scope of structures, systems, and components (SSCs) requiring special treatment. ''Special treatment requirements'' refers to those specific examples of regulations that are applied in order to provide a high degree of assurance that SSC will be capable of performing their intended functions when needed. The current scope of SSCs covered by the special treatment requirements governing commercial nuclear reactors is deterministically based and stems primarily from the evaluation of selected design basis events, as described in updated final safety analysis reports (UFSARs). This regulatory framework provides reasonable assurance of no undue risk to the health and safety of the public. However, recent advances in technology, coupled with operating reactor experience, have suggested that an alternative approach that would use a risk-informed process for evaluating SSC safety significance, would, in turn, result in a more focused determination of which SSCs should receive special treatment requirements. (author)

Completely automated nuclear reactors for long-term operation

International Nuclear Information System (INIS)

Teller, E.; Ishikawa, M.; Wood, L.

1996-01-01

The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

Operational methods of the fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Borges, V.; Sefidvash, F.

1993-01-01

The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

Review of Operation and Maintenance Support Systems for Research Reactors

International Nuclear Information System (INIS)

Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

2014-01-01

Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

Fuel management for TRIGA reactor operators

International Nuclear Information System (INIS)

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

1980-01-01

One responsibility of the Supervisor of Reactor Operations is to follow the TRIGA core depletion and recommend core loading changes for refueling and special experiments. Calculations required to analyze such changes normally use digital computers and are extremely difficult to perform for one who is not familiar with computer language and nuclear reactor diffusion theory codes. The TRICOM/SCRAM program developed to perform such calculations for the Penn State TRIGA Breazeale Reactor (PSBR), has a very simple input format and is one which can be used by persons having no knowledge of computer codes. The person running the program need not understand computer language such as Fortran, but should be familiar with reactor core geometry and effects of loading changes. To further simplify the input requirements but still allow for all of the studies normally needed by the reactor operations supervisor, the options required for input have been isolated to two. Given a master deck of computer cards one needs to change only three cards; a title card, core energy history information card and one with core changes. With this input, the program can provide individual fuel element burn-up for a given period of operation and the k eff of the core. If a new loading is desired, a new master deck containing the changes is also automatically provided. The life of a new core loading can be estimated by feeding in projected core burn-up factors and observing the resulting loss in individual fuel elements. The code input and output formats have now been made sufficiently convenient and informative as to be incorporated into a standard activity for the Reactor Operations Supervisor. (author)

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

Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

International Nuclear Information System (INIS)

Whitehead, D.W.; Hake, T.M.

1990-01-01

Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

Risk Management for Sodium Fast Reactors.

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

International Nuclear Information System (INIS)

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

1984-12-01

During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

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

International Nuclear Information System (INIS)

1985-04-01

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

Tendencies in operating power reactors

International Nuclear Information System (INIS)

Brinckmann, H.F.

1987-01-01

A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

Safe operation of critical assemblies and research reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1960-09-15

Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

International Nuclear Information System (INIS)

Gotschall, H.L.

1994-01-01

The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

Simulation for Remote Operation for REX10 Nuclear Reactor

International Nuclear Information System (INIS)

Lee, Sim Won; Kim, Dong Su; Na, Man Gyun; Lee, Yoon Joon; Lee, Yeon Gun; Park, Goon Cherl

2010-01-01

The newly designed REX10 (Regional Energy Reactor, 10MWth) is an environmentally-friendly and stable small nuclear reactor for a small-scale reactor based Multi-purpose regional energy system. The REX10 has been developed to maintain system safety in order to be placed in densely populated region, island, etc. In addition, it is significantly hard to recruit many operation and maintenance personnel for small power reactors differently from usual commercial reactors because of its remote location and of economic reasons. In order to overcome these constraints, to decrease the operation cost by reducing operation and maintenance personnel, and to increase plant reliability through autonomous plant control, it is needed to design the control system of the small power reactors and to establish its unmanned remote operation system. In this study, the REX10 reactor core thermal power controller is designed by using a REX10 code analyzer. The remote control facility through man-machine interface (MMI) design and interface between programming languages was established and it was used to verify remote operation of REX10

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Small reactor operating mode

International Nuclear Information System (INIS)

Snell, V.G.

1997-01-01

There is a potential need for small reactors in the future for applications such as district heating, electricity production at remote sites, and desalination. Nuclear power can provide these at low cost and with insignificant pollution. The economies required by the small scale application, and/or the remote location, require a review of the size and location of the operating staff. Current concepts range all the way from reactors which are fully automatic, and need no local attention for days or weeks, to those with reduced local staff. In general the less dependent a reactor is on local human intervention, the greater its dependence on intrinsic safety features such as passive decay heat removal, low-stored energy and limited reactivity speed and depth in the control systems. A case study of the design and licensing of the SLOWPOKE Energy System heating reactor is presented. (author)

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

1994-01-01

The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

Operating manual for the Health Physics Research Reactor

International Nuclear Information System (INIS)

1985-11-01

This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

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

Reactor operation method

International Nuclear Information System (INIS)

Suzuki, Toshio; Hida, Kazuki; Yoshioka, Ritsuo.

1990-01-01

The enrichment degree of fuels initially loaded in a reactor core was made extremely lower than that of fresh fuels to be loaded in the succeeding cycle, or the enrichment degree for all of the initially loaded fuels was made identical with that of the fresh fuels in the conventional reactor operation method. In this operation method, since the initially loaded fuels are sometimes taken out after the completion of the cycle at the low burnup degree as it is, it can not be said to reduce the fuel cycle cost. As a means for dissolving this problem, at least two different kinds of initially loaded fuels are prepared. The enrichment degree of the highly enriched fuels is made identical with that of the fresh fuels, and the enrichment degree and the number of low enriched fuels are not changed after the completion of the first cycle but they are operated till the end of the second cycle. Further, all of the fuels at the low enrichment degree are taken out after the completion of the second cycle and exchanged with the fresh fuels. As a result, high burnup ratio of the initially loaded fuels can be increased, to improve the fuel economy. (I.S.)

Method of operating BWR type reactors

International Nuclear Information System (INIS)

Sekimizu, Koichi

1980-01-01

Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

Operating manual for the Bulk Shielding Reactor

International Nuclear Information System (INIS)

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

Operating manual for the Bulk Shielding Reactor

International Nuclear Information System (INIS)

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

Operating manual for the Bulk Shielding Reactor

Energy Technology Data Exchange (ETDEWEB)

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

Operating manual for the Bulk Shielding Reactor

Energy Technology Data Exchange (ETDEWEB)

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

Artificial intelligence in nuclear reactor operation

International Nuclear Information System (INIS)

Da Ruan; Benitez-Read, J.S.

2005-01-01

Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

Manual for the operation of research reactors

International Nuclear Information System (INIS)

1965-01-01

The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

Safety of research reactors (Design and Operation)

International Nuclear Information System (INIS)

Dirar, H. M.

2012-06-01

The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

Principle of an operational complexity index for the characterization of the human factor relevance of future reactors concepts

International Nuclear Information System (INIS)

Papin, Bernard

2004-01-01

With the increasing reliability of the modern technological systems, the human contribution to the global risk in the operation of industrial systems is becoming more and more significant : in the nuclear reactor operation for example, a recent PSA estimation of this contribution is about 25% of the risk of core melting, all situations considered. This urges the designers of future nuclear reactors to consider the minimisation of this Human Factor (HF) contribution, at the very early stage of their design : the experience feedback shows that this is indeed at this stage that the fundamental design options, impacting the most the human reliability in operation, are fixed. The problem is that at these early design stages, it is also quite impossible to apply formal human reliability methods to support this HF optimisation, while the precise operating conditions of the reactor are not yet known in enough details. In this paper, another approach of the HF evaluation during the design, based on the functional and operational complexity assessment, is proposed. As an illustration, this approach is used to compare various concepts of Pressurized Water Reactors from the point of view of the Human Factor relevance. (Author)

RA reactor operation and maintenance in 1994, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

1994-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

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

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

Regulations for RA reactor operation; Propisi nuklearnog reaktora 'RA'

Energy Technology Data Exchange (ETDEWEB)

NONE

1980-09-15

Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions. [Serbo-Croat] Propisi o radu nuklearnog reaktora RA pisani su tako da svi zakonski propisi definisani 'Zakonom o zastiti od jonizujuceg zracenja' i pratecim propisima (devet pravilnika) kao i tehnicke norme prema preporukama MAAE budu postovani u punoj meri pri radu reaktora. Sadrzaj ove knjige obuhvata: osnovne podatke o reaktoru; zakonske propise; organizaciju rada reaktora RA; opste propise o rezimu rada, kretanju u zgradi reaktora, izvodjenju eksperimenata; pogonske propise za rad u normalnom rezimu i u slucaju udesa.

Operation and utilization of Indonesia Research Reactors

International Nuclear Information System (INIS)

Kuntoro, Iman; Sujalmo, Saiful; Tarigan, Alim

2004-01-01

For supporting the R and D in nuclear science and technology and its application, BATAN own and operate three research reactors namely, TRIGA-2000, KARTINI and RSG-GAS having thermal power of 2 MW, 100 kW and 30 MW respectively. The main features, operation and utilization progress of the reactors are described in this report. (author)

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Operating reactors licensing actions summary. Volume 5, No. 6

International Nuclear Information System (INIS)

1985-08-01

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

Operational safety evaluation for minor reactor accidents

International Nuclear Information System (INIS)

Wang, O.S.

1981-01-01

The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

Component Degradation Susceptibilities As The Bases For Modeling Reactor Aging Risk

International Nuclear Information System (INIS)

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2010-01-01

The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1985-01-01

This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

Evaluating the Cost, Safety, and Proliferation Risks of Small Floating Nuclear Reactors.

Science.gov (United States)

Ford, Michael J; Abdulla, Ahmed; Morgan, M Granger

2017-11-01

It is hard to see how our energy system can be decarbonized if the world abandons nuclear power, but equally hard to introduce the technology in nonnuclear energy states. This is especially true in countries with limited technical, institutional, and regulatory capabilities, where safety and proliferation concerns are acute. Given the need to achieve serious emissions mitigation by mid-century, and the multidecadal effort required to develop robust nuclear governance institutions, we must look to other models that might facilitate nuclear plant deployment while mitigating the technology's risks. One such deployment paradigm is the build-own-operate-return model. Because returning small land-based reactors containing spent fuel is infeasible, we evaluate the cost, safety, and proliferation risks of a system in which small modular reactors are manufactured in a factory, and then deployed to a customer nation on a floating platform. This floating small modular reactor would be owned and operated by a single entity and returned unopened to the developed state for refueling. We developed a decision model that allows for a comparison of floating and land-based alternatives considering key International Atomic Energy Agency plant-siting criteria. Abandoning onsite refueling is beneficial, and floating reactors built in a central facility can potentially reduce the risk of cost overruns and the consequences of accidents. However, if the floating platform must be built to military-grade specifications, then the cost would be much higher than a land-based system. The analysis tool presented is flexible, and can assist planners in determining the scope of risks and uncertainty associated with different deployment options. © 2017 Society for Risk Analysis.

The development of reactor operator license examination question bank

International Nuclear Information System (INIS)

Kim, In Hwan; Woo, S. M.; Kam, S. C.; Nam, K. J.; Lim, H. P.

2001-12-01

The number of NPP keeps increasing therefore there is more need of reactor operators. This trend requires the more efficiency in managing the license examination. Question bank system will help us to develop good quality examination materials and keep them in it. The ultimate purpose of the bank system is for selecting qualified reactor operators who are primarily responsible for the safety of reactor operation in NPP

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Control rod for the operation of nuclear reactor

International Nuclear Information System (INIS)

Ishida, Hiromi

1987-01-01

Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

Report of the reactor Operators Service - Annex F

International Nuclear Information System (INIS)

Zivotic, Z.

1990-01-01

RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. During 1989 the operators service staff participated in the following activities: reconstruction of the existing reactor systems, control of the emergency cooling system, construction of the experimental loop 'Vinca-1'. Education of the staff was organized through routine courses, practical training is foreseen for 1991 [sr

Licensed operating reactors

International Nuclear Information System (INIS)

Hartfield, R.A.

1994-03-01

The Nuclear Regulatory Commissions annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December, the year to date (in this case calendar year 1993) and cumulative data, usually for the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

Operation monitoring and protection method for nuclear reactor

International Nuclear Information System (INIS)

Tochihara, Hiroshi.

1995-01-01

In an operation and monitoring method for a PWR-type reactor by using a tetra-sected neutron detector, axial off set is defined by neutron detector signals with respect to an average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core. A departure from nucleate boiling (DNBR) is represented by standardized signals, and the DNBR is calculated by using the axial off set of the average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core, and they are graphically displayed. In addition, a thermal flow rate-water channel coefficient is also graphically displayed, and the DNBR and the thermal flow rate-water channel coefficient are restricted based on the display, to determine an allowable operation range. As a result, it is possible to provide an operation monitoring and protection method for nuclear reactor capable of reducing labors and frequencies for the change of protection system setting in a case of using a tetra-sected neutron detector disposed at the outside and, at the same time, protecting each of DNR and the highest linear power or the thermal water coefficient channel. (N.H.)

Risk assessment of computer-controlled safety systems for fusion reactors

International Nuclear Information System (INIS)

Fryer, M.O.; Bruske, S.Z.

1983-01-01

The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Science.gov (United States)

2013-11-29

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0260] Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this Federal Register notice to inform the public of a slight change in the manner of distribution of publicly available operating reactor licensing...

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

Energy Technology Data Exchange (ETDEWEB)

Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z., E-mail: raissaomarques@gmail.com, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: silvasf@cdtn.br, E-mail: amandaraso@hotmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

Probabilistic safety analysis and risk-based inspection of nuclear research reactors: state-of-the-art and implementation proposal

International Nuclear Information System (INIS)

Marques, Raíssa O.; Vasceoncelos, Vanderley de; Soares, Wellington A.; Silva Júnior, Silvério F.; Raso, Amanda L.; Mesquita, Amir Z.

2017-01-01

Industrial facilities systems deteriorate over time during operation, thus increasing the possibility of accidents. Risk-Based Inspection (RBI) classifies such systems by their risk information with the purpose of prioritizing inspection efforts. RBI can reduce inspection activities, resulting in lower risk levels, and maintaining reliability and safety in acceptable levels. Risk-Informed In-Service Inspection (RI-ISI) is a RBI approach used in nuclear industry. RI-ISI uses outcomes from Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPP) to plan In-Service Inspections (ISI). Despite nuclear research reactors are simpler and have lower risks than power reactors, the application of PSA to them may be useful for safety improvements once they are more flexible, provide easier access to its core, and allow changes in fuel configurations in case of experimental tests. Ageing management of structures, systems and components important to safety of a nuclear research reactor throughout its lifetime is also required to assure continued adequacy of safety levels, reliable operation, and compliance with operational limits and conditions. This includes periodic review of ISI programs in which monitoring of material deterioration and aging effects are considered, and that can be supported by the RBI approach. A review of state-of-the-art of PSA and RBI applications to nuclear reactors is presented in this work. Advantages to apply these methodologies are also analyzed. PSA and RBI implementation proposal applied to nuclear research reactors is also presented, as well as its application to a TRIGA research nuclear reactor using computer codes developed by ReliaSoft® Corporation. (author)

Method of controlling the reactor operation

International Nuclear Information System (INIS)

Ishiguro, Akira; Nakakura, Hiroyuki.

1987-01-01

Purpose: To moderate vibratory response due to delayed operation thereby obtain stable controlled response in the operation control for a PWR type reactor. Method: the reactor operation is controlled by the axial power distribution control by regulating the boron concentration in primary coolants with a boron density control system and controlling the average temperature for the primary coolants with the control rod control system. In this case, the control operation and the control response become instable due to transmission delay, etc. of aqueous boric acid injection to the primary coolant circuits to result in vibratory response. In the present invention, signals are prepared by adding the amount in proportion to the variation coefficient with time of xenone concentration obtained from the measured value for the reactor power added to the conventional axial power distribution parameter deviation and used as the input signals for the boron concentration control system. As a result, the instability due to the transmission delay of the aqueous boric acid injection is improved by the preceding control by the amount in proportion with the variation coefficient with time of the xenone concentration. An advantageous effect can be expected for the load following operation during day time according to the present invention. (Kamimura, M.)

RA reactor operation and maintenance in 1996, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

1996-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

International Nuclear Information System (INIS)

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

1988-01-01

According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

14th Biennial conference on reactor operating experience plant operations: The human element

International Nuclear Information System (INIS)

Anon.

1989-01-01

Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

LBB application in the US operating and advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

Operating history of U.S. nuclear power reactors

International Nuclear Information System (INIS)

1974-01-01

The operating history of U. S. nuclear power plants through December 31, 1974 has been collected. Included are those nuclear reactor facilities which produce electricity, even if in token amounts, or which are part of a development program concerned with the generation of electricity through the use of a nuclear reactor as a heat source. The information is based on data furnished by facility operators. The charts are plotted in terms of cumulative thermal energy as a function of time. Since only those shutdowns of five days or more are shown, the charts do not give a detailed history of plant operation. They do, however, give an overview of the operating history of a variety of developmental and experimental nuclear power reactors. The data show the yearly gross generation of electricity for each U. S. nuclear plant and, for civilian power plants, information on reactor availability and plant capacity factor. (U.S.)

Licensed operating reactors

International Nuclear Information System (INIS)

1991-08-01

The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar 1990) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

Sustainability management for operating organizations of research reactors

Energy Technology Data Exchange (ETDEWEB)

Kibrit, Eduardo; Aquino, Afonso Rodrigues de, E-mail: ekibrit@ipen.br, E-mail: araquino@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

Sustainability management for operating organizations of research reactors

International Nuclear Information System (INIS)

Kibrit, Eduardo; Aquino, Afonso Rodrigues de

2017-01-01

Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

Nuclear safety requirements for operation licensing of Egyptian research reactors

International Nuclear Information System (INIS)

Ahmed, E.E.M.; Rahman, F.A.

2000-01-01

From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

Operating reactors licensing actions summary. Vol.4, No. 4

International Nuclear Information System (INIS)

1984-06-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

Operational behaviour of a reactor normal operation and disturbances

International Nuclear Information System (INIS)

Geyer, K.H.

1982-01-01

During normal operation, the following topics are dealt with: primary and secondary coolant circuits - full load operation - start-up and shutdown - steady state part load diagramm. During disturbances and incidents, the following procedures are discussed: identification and detection of the events - automatic actions - manual actions of the operator - provided indications - explanation of actuated systems - basic information of reactor protection system. (RW)

Operation of the OSIRIS reactor from the viewpoint of analysis of operator functions

International Nuclear Information System (INIS)

Fichet-Clairfontaine, P.Y.; Saint-Jean, T.

1985-09-01

This paper presents the results of analyses carried out on site by the Human Factor Study Laboratory in an experimental nuclear plant operated by the Atomic Energy Commissariat - the OSIRIS pool reactor. The analyses of certain tasks are given: work in the reactor hall and an operation of circuit setting performed by the mechanics. This work has thrown light on certain operational guidelines implemented by the operators when carrying out their work [fr

Research about reactor operator's personability characteristics and performance

Energy Technology Data Exchange (ETDEWEB)

Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

2003-03-01

To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

Operating reactors licensing actions summary. Volume 5, No. 7

International Nuclear Information System (INIS)

1985-09-01

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

Upgrade of reactor operation technology

International Nuclear Information System (INIS)

Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

2003-01-01

To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

Operation management of the prototype heavy water reactor 'Fugen'

Energy Technology Data Exchange (ETDEWEB)

Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

1983-09-01

The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported.

Reactor operation environmental information document

Energy Technology Data Exchange (ETDEWEB)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Independent verification: operational phase liquid metal breeder reactors

International Nuclear Information System (INIS)

Bourne, P.B.

1981-01-01

The Fast Flux Test Facility (FFTF) recently achieved 100-percent power and now is in the initial stages of operation as a test reactor. An independent verification program has been established to assist in maintaining stable plant conditions, and to assure the safe operation of the reactor. Independent verification begins with the development of administrative procedures to control all other procedures and changes to the plant configurations. The technical content of the controlling procedures is subject to independent verification. The actual accomplishment of test procedures and operational maneuvers is witnessed by personnel not responsible for operating the plant. Off-normal events are analyzed, problem reports from other operating reactors are evaluated, and these results are used to improve on-line performance. Audits are used to confirm compliance with established practices and to identify areas where individual performance can be improved

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

L-Reactor operation, Savannah River Plant: environmental assessment

International Nuclear Information System (INIS)

1982-08-01

The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

Small nuclear reactor safety design requirements for autonomous operation

International Nuclear Information System (INIS)

Kozier, K.S.; Kupca, S.

1997-01-01

Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Understanding to requirements for educational level in qualification of reactor operators

International Nuclear Information System (INIS)

Zhang Chi; Yang Di; Zhou Limin

2007-01-01

Requirements for qualification of reactor operators in nuclear safety regulations were discussed in this paper. The new issue was described in the confirmation of education level of reactor operators. The understanding to the requirements for Educational Level in Qualification of Reactor Operators was provided according to Higher Education Law of the People's Republic of China. It was proposed to improve the confirmation of qualification of reactor operators as soon as possible. (authors)

Procedures to relate the NII safety assessment principles for nuclear reactors to risk

CERN Document Server

Kelly, G N; Hemming, C R

1985-01-01

Within the framework of the Public Inquiry into the proposed pressurised water reactor (PWR) at Sizewell, estimates were made of the levels of individual and societal risk from a PWR designed in a manner which would conform to the safety assessment principles formulated by the Nuclear Installations Inspectorate (NII). The procedures used to derive these levels of risk are described in this report. The opportunity has also been taken to revise the risk estimates made at the time of the Inquiry by taking account of additional data which were not then available, and to provide further quantification of the likely range of uncertainty in the predictions. This re-analysis has led to small changes in the levels of risk previously evaluated, but these are not sufficient to affect the broad conclusions reached before. For a reactor just conforming to the NII safety assessment principles a maximum individual risk of fatal cancer of about 10 sup - sup 6 per year of reactor operation has been estimated; the societal ris...

RA reactor operation and maintenance in 1989, Part 1

International Nuclear Information System (INIS)

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

1989-01-01

During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

Strategies of operation cycles in BWR type reactors

International Nuclear Information System (INIS)

Molina, D.; Sendino, F.

1996-01-01

The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

Analysis of the radiometric survey during the Argonauta reactor operation

International Nuclear Information System (INIS)

Oliveira, Eara de S.L.; Cardozo, Katia K.M.; Silva, Joao Carlos P.; Santos, Joao Regis dos

2013-01-01

The Argonaut reactor at the Institute of Nuclear Engineering-IEN/CNEN, operates normally, the powers between 1.7 and 340 W on neutrongraphy procedures, production of radionuclides and experimental reactor physics lessons to postgraduate courses. The doses from neutrons and gamma radiation are measured when the reactor is critical, inside the reactor hall and surrounding regions. A study of the data obtained was performed to evaluate the daily need of this survey in the reactor hall. Taking into account the principle ALARA, which aims to optimize and minimize the dose received by the individual, we propose, in this work, through an analysis of the acquired data in occupational radiometric surveys, a reformulation of the area monitoring routine practiced by the team of radiological protection of the Institute of Nuclear Engineering - IEN/CNEN-RJ, whereas other monitoring routines regarding the radiological protection are also applied in the routine of the reactor. The operations under review occurred with the reactor operating 340 W power at intervals of 60, 120 and 180 minutes, in monitoring points in controlled areas, supervised and free. The results showed significant dose values in the output of the J-Channel 9 when the operation occurs with this open. With 180 minutes of operation, the measured values of dose rate were lower than the values at 60 min and 120 operations min. At the point in the supervised area, offsite to the reactor hall, situated in the direction of the J-Channel 9, the value reduces more than 14% in any operating time in relation to the dose rate measured at the point opposite the canal. There is a 50% reduction in the dose rates for operations with and J-9 closed. The results suggest a new frequency of radiometric survey whose mode of operation is maintained in similar conditions, since combined with other relevant practices of radiation protection

Research nuclear reactor RA - Annual Report 1975. Operation and maintenance

International Nuclear Information System (INIS)

Martinc, R.

1976-01-01

The plan for 1975 was successfully fulfilled. This is reflected in research related to improvement of operating properties of the RA reactor, mostly due to the effort of the RA staff employed in operation and maintenance of the reactor. Fuel saving achieved by this activity amounted to about 38% (80% enriched fuel). Preliminary work is done, concerned with new reactor core with highly enriched fuel. This is a significant saving as well. New fuel elements have arrived at the end of this year. It is going to enable increase of neutron flux by 50% without changing the nominal operating power. The possibility of further improvement of the reactor are analyzed, to enable material testing and production of radioactive sources. Mid term plan for reactor operation was made according to this analysis. It is planned to further increase the neutron flux in isolated smaller zones, and building new experimental loops with cooling and fast neutron converters. Much was done to increase the safety level of reactor operation and preparing the safety report [sr

Method of operating FBR type reactors

International Nuclear Information System (INIS)

Arie, Kazuo.

1984-01-01

Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

Risk-assessment methodology for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.

1976-04-01

The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

Reactor operating procedures for start up of continuously operated chemical plants

NARCIS (Netherlands)

Verwijs, J.W.; Verwijs, J.W.; Kösters, P.H.; van den Berg, Henderikus; Westerterp, K.R.; Kosters, P.G.H.

1995-01-01

Rules are presented for the startup of an adiabatic tubular reactor, based on a qualitative analysis of the dynamic behavior of continuously-operated vapor- and liquid-phase processes. The relationships between the process dynamics, operating criteria, and operating constraints are investigated,

1984 Operation of the high flux reactor

International Nuclear Information System (INIS)

1985-01-01

The programme resources in 1984 were largely devoted to the replacement of the old reactor vessel and its peripheral equipment. The original vessel had been in operation for more than 20 years and doubts had arisen about the condition of the aluminium tank after so long an exposure to neutrons. The operation, which had never been attempted before on a reactor of that size and complexity was planned and prepared over a number of years to take advantage of the occasion to provide a much improved vessel, incorporating the latest design features. The plant was shut down at the end of November 1983 and the 14 months operation began with a short cooling-off period for decay of short lived radioactivity followed by removal of the old tank and its dissection into pieces convenient for consolidation and storage as radioactive waste. After decontamination of the shielding pool, the new vessel and neutron beam tubes were installed and the reactor was recommissioned. Routine 45 MW operation was resumed on 14 February 1985 and has been uneventful since then

Operation experience with the TRIGA reactor Wien 2004

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-01-01

The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

Annual report 1990. Operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1990-01-01

In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

Annual report 1989 operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1989-01-01

In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

Operating experience with the DRAGON High Temperature Reactor experiment

International Nuclear Information System (INIS)

Simon, R.A.; Capp, P.D.

2002-01-01

The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

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

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

Operation and utilizations of Dalat nuclear research reactor

International Nuclear Information System (INIS)

Hien, P.Z.

1988-01-01

The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

Operating reactors licensing actions summary. Volume 5, Number 1

International Nuclear Information System (INIS)

1985-03-01

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

General areas needing chemical competence to support reactor operation

International Nuclear Information System (INIS)

Proksch, E.; Bildstein, H.

1963-01-01

Chemical competence is needed not only for the development of new types of reactors but also for the start-up and safe operation of reactors. The activities of chemistry and chemical engineering cover a number of fields, namely chemical analysis, radiochemical analysis, corrosion research, radiolysis of water and water purification. The author reviews fields in reactor operation and maintenance in which chemical competence is needed. (author). 9 refs

The evaluation of operator reliability factors on power reactor

International Nuclear Information System (INIS)

Karlina, Itjeu; Supriatna, Piping; W, Suharyo; Santosa, Kussigit; Darlis; S, Bambang; Y, Sasongko

1999-01-01

The sophisticated technology system was not assured the reliability system itself because it has contained a part of human dependence affected successfully of reactor operation either how work smoothly and safe or failure ac cured and then accident appears promptly. The evaluation of operator reliability factor on ABWR power reactor has been carried out which consist of criterion skill and workload according to NUREG/CR-2254, NUREG/CR-4016 and NUREG-0835 the reactor operation reliability emphasize to the operator are synergic between skill and workload themselves. The employee's skill will affect to the type and level of their tasks. The operator's skill depend on education and experiences, position or responsibility of tasks, physical conditions (age uninvalid of physic/mental

Neutron field control cybernetics model of RBMK reactor operator

International Nuclear Information System (INIS)

Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

1992-01-01

Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

Reactor modification, preparation and operation

International Nuclear Information System (INIS)

Weill, J.; Furet, J.; Baillet, J.; Donvez, G.; Duchene, J.; Gras, R.; Mercier, R.; Chenouard, J.; Leconte, J.

1962-01-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-01

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

Reactor modification, preparation and operation

Energy Technology Data Exchange (ETDEWEB)

Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

1962-03-15

In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

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

RA reactor operation and maintenance in 1992, Part 1

International Nuclear Information System (INIS)

Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

1992-01-01

During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

Management of operational events in research reactor

International Nuclear Information System (INIS)

Zhong Heping; Yang Shuchun; Peng Xueming

2001-01-01

The author describes the tracing management process post-operational event in a research reactor based on nuclear safety code, under the background of the research reactor in Nuclear Power Institute of China. It presorts the definite measures to the event tracing and it up its management factors

Fluidized-bed reactors processes and operating conditions

CERN Document Server

Yates, John G

2016-01-01

The fluidized-bed reactor is the centerpiece of industrial fluidization processes. This book focuses on the design and operation of fluidized beds in many different industrial processes, emphasizing the rationale for choosing fluidized beds for each particular process. The book starts with a brief history of fluidization from its inception in the 1940’s. The authors present both the fluid dynamics of gas-solid fluidized beds and the extensive experimental studies of operating systems and they set them in the context of operating processes that use fluid-bed reactors. Chemical engineering students and postdocs as well as practicing engineers will find great interest in this book.

Reactor operator screening test experiences

International Nuclear Information System (INIS)

O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

1976-01-01

When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

ENSI's view on technical safety for the long term operation of reactors 1 and 2 in the Beznau nuclear power plant

International Nuclear Information System (INIS)

2010-11-01

The reactors 1 and 2 of the Beznau nuclear power plant (KKB) are operated since about 40 years. For an operation beyond the design period of 40 years the Swiss Federal Nuclear Safety Inspectorate (ENSI) demands the evidence to be brought that the design limits of the safety relevant components will not be reached during the extended operation period. In 2008 the license holder of KKB delivered the requested documentation on material ageing on the basis of deterministic as well as probabilistic safety analyses and concluded that both reactors can be safely operated beyond 40 years. Thanks to continuous additional outfits, both reactors are in good condition from the point of view of technical safety. With a view to the extension of operation beyond 40 years, KKB already applied the necessary measures regarding technics, finances and personnel in order to keep the present technical level. Since 1991 KKB has analysed and checked components that are difficult to replace. From the evidence presented, ENSI concluded that both reactors are able to be operated up to 60 years long, however with two restrictions for reactor 1 because there the material used for the reactor pressure vessel (RPV) suffered more neutron brittleness than in reactor 2. In addition, reactor 1 is much more affected by ageing phenomena than reactor 2, but, according to neutron fluence calculations, the limiting criteria will not be reached even after 60 years of operation. Some corrosion damages were noted at the lower part of the RPV due to water containing boron acid; they are more pronounced in reactor 1 than in reactor 2. Even though the calculations done by KKB are very conservative, they show that also in the long term the operation limiting criteria about the mechanical resistance of the RPV are never reached. ENSI concludes that the safety design of both KKB reactors ensures safe control of the design basis accidents. Both reactors were continuously fitted with new equipment. With the planed

A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

Energy Technology Data Exchange (ETDEWEB)

Jacques V Hugo; David I Gertman; Jeffrey C Joe

2014-08-01

This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

Research Project 'RB research nuclear reactor' (operation and maintenance), Final report

International Nuclear Information System (INIS)

1985-01-01

This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training

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

Annual Report 1991. Operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1992-01-01

In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

Mode of operation of a nuclear reactor

International Nuclear Information System (INIS)

Morita, T.

1976-01-01

A method is proposed for the operation of a nuclear reactor guaranteeing an essentially symmetrical axial power distribution during normal operation by controlling the changes occuring in the reactor power partly by variation of the concentration of the neutron-absorbing element and partly by variation of the control rod positions. The representative parameters are recorded for the upper and lower half and adjusted to a predetermined reference value. In using this method, the axial power peals are reduced and power losses avoided. (RW) [de

RA reactor operation and maintenance

International Nuclear Information System (INIS)

Zecevic, V.

1963-02-01

This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

Operational experience with Dragon reactor experiment of relevance to commercial reactors

International Nuclear Information System (INIS)

Capp, P.D.; Simon, R.A.

1976-01-01

An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

Method of operating water cooled reactor with blanket

International Nuclear Information System (INIS)

Suzuki, Katsuo.

1988-01-01

Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

Regulations and instructions for RB reactor operation; Propisi i uputstva za rad reaktora RB

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-07-01

This document includes regulations for reactor RB operation, behaviour and presence of staff in the reactor building; regulations for performing experiments at the RB reactor, regulations and int ructions for the reactor operators and other staff on duty. A chapter is devoted to instruction for reactor operation with the operating documentation and special duties of the operators. Regulations and instruction concerned with accidents are described with classification of accidents and evacuation plan. Annexes to this document include: the present status of the reactor; program for training the reactor operators; forms which are obligatory to be signed for any operating activity, and the certificate of the RB reactor lattice.

Operation, maintenance and utilization of the RA reactor, Annual report 1978

International Nuclear Information System (INIS)

Milosevic, M.

1978-12-01

It has been planned for 1978 that the RA reactor would be operated for 158 dana at nominal power of 6.5 MW meaning production of 24 648 MWh. The plan was fulfilled since 24 652 MWh was produces. Reactor operation for 158 days is relevant to reactor operation for 200 days in the period before 1975. The reason is increased neutron flux achieved due to improved fuel management and the characteristics of the new 80% enriched fuel. At the end of 1978 the reactor core contained 45% of 80% enriched fuel elements. Increase of neutron flux has shortened the typical time needed for irradiation of the most important samples for isotope production. This significant success in reactor operation is at the same time an obligation for increasing its utilization. Some new trends proposed for increasing reactor utilization capacities were presented at the Conference on utilization of research nuclear reactors in Yugoslavia held in May 1978 [sr

Operation of Packed-Bed Reactors Studied in Microgravity

Science.gov (United States)

Motil, Brian J.; Balakotaiah, Vemuri

2004-01-01

The operation of a packed bed reactor (PBR) involves gas and liquid flowing simultaneously through a fixed-bed of solid particles. Depending on the application, the particles can be various shapes and sizes but are generally designed to force the two fluid phases through a tortuous route of narrow channels connecting the interstitial space. The PBR is the most common type of reactor in industry because it provides for intimate contact and high rates of transport between the phases needed to sustain chemical or biological reactions. The packing may also serve as either a catalyst or as a support for growing biological material. Furthermore, this type of reactor is relatively compact and requires minimal power to operate. This makes it an excellent candidate for unit operations in support of long-duration human space activities.

Fractional power operation of tokamak reactors

International Nuclear Information System (INIS)

Mau, T.K.; Vold, E.L.; Conn, R.W.

1986-01-01

Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

International Nuclear Information System (INIS)

Rokhmadi

2007-01-01

Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

Extending the maximum operation time of the MNSR reactor.

Science.gov (United States)

Dawahra, S; Khattab, K; Saba, G

2016-09-01

An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

Procedure for operating a heavy water cooled power reactor

International Nuclear Information System (INIS)

Rau, P.; Kumpf, H.

1981-01-01

Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

Graphics and control for in-reactor operations

International Nuclear Information System (INIS)

Smith, A.L.

1996-01-01

A wide range of manipulator systems has been developed to carry out remotely operated inspection, repair and maintenance tasks at the Magnox reactors in the United Kingdom. A key factor in the improvement of these systems in recent years has been the extensive use of computer graphics as a real-time aid to the manipulator operator. This is exemplified by the reactor pressure vessel inspection work at the Bradwell reactor which is described in detail. The graphics sub-system of the control system for the manipulator plays a unique and wide-ranging role. The 3D modelling and simulation capability of the IGRIP software has contributed to the conceptual design, detailed path planning, rehearsal support, public relations, real-time manipulator display, post inspection documentation and quality assurance. (UK)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Systemic model for the aid for operating of the reactor Siloe

International Nuclear Information System (INIS)

Royer, J.C.; Moulin, V.; Monge, F.

1995-01-01

The Service of the Reactor Siloe (CEA/DRN/DRE/SRS), fully aware of the abilities and knowledge of his teams in the field of research reactor operating, has undertaken a project of knowledge engineering in this domain. The following aims have been defined: knowledge capitalization for the installation in order to insure its perenniality and valorization, elaboration of a project for the aid of the reactor operators. This article deals with the different actions by the SRS to reach the aims: realization of a technical model for the operation of the Siloe reactor, development of a knowledge-based system for the aid for operating. These actions based on a knowledge engineering methodology, SAGACE, and using industrial tools will lead to an amelioration of the security and the operating of the Siloe reactor. (authors). 13 refs., 7 figs

Development of operator thinking model and its application to nuclear reactor plant operation system

International Nuclear Information System (INIS)

Miki, Tetsushi; Endou, Akira; Himeno, Yoshiaki

1992-01-01

At first, this paper presents the developing method of an operator thinking model and the outline of the developed model. In next, it describes the nuclear reactor plant operation system which has been developed based on this model. Finally, it has been confirmed that the method described in this paper is very effective in order to construct expert systems which replace the reactor operator's role with AI (artificial intelligence) systems. (author)

The operating reliability of the reactor coolant pump

International Nuclear Information System (INIS)

Grancy, W.

1996-01-01

There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

Artificial intelligence program in a computer application supporting reactor operations

International Nuclear Information System (INIS)

Stratton, R.C.; Town, G.G.

1985-01-01

Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II

Code on the safety of nuclear research reactors: Operation

International Nuclear Information System (INIS)

1992-01-01

The purpose of this publication is to provide the essential requirements and recommendations for the safe operation of research reactors, with emphasis on the supervisory and managerial aspects. However, the publication also provides some guidance and information on topics concerning all the organizations involved in operation. These objectives are expressed in terms of requirements and recommendations for the safe operation of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on the ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop regulations and safety criteria for its research reactor programme.

The risks of nuclear energy technology. Safety concepts of light water reactors

Energy Technology Data Exchange (ETDEWEB)

Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

2014-11-01

Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

Dynamics of nuclear reactor operational cycles

International Nuclear Information System (INIS)

Chapman, L.D.; Wayland, J.R.

With this system dynamics computer model, one can explore the long term effects of a nuclear reactor program. Given an input demand for reactors, the consequences on each sector and the interactions among sectors can be simulated to provide a better understanding of the time development of a nuclear reactor program. The model permits the determination of various levels of activity as a function of time for plant enrichment, fuel fabrication, fuel reprocessing and storage of waste products. In addition, the rates of construction of reactors, spent fuel transit, disposal of waste, mining, shipping, recycling and enrichment can be investigated for optimal planning purposes. The model has been written in a very general manner so that it can be used to simulate any nuclear reactor program. It is an easy task to relate the amount of accidental or operational release of radioactive contaminants into our environment to the activity levels of each of the above sectors. (U.S.)

Systems for aiding operators at university-owned research reactors in Japan

International Nuclear Information System (INIS)

Nishihara, H.; Kimura, Y.; Shibata, T.

1984-01-01

University-owned research reactors are operated for various purposes, and small disturbances may arise from various experimental facilities. Also not uniform are the technical levels of operators who range from supervised-students to reactor physicists. Considerable efforts are therefore devoted to the preventive maintainance. With these boundary conditions imposed, systems for aiding operators are designed at these research reactor facilities. (author)

Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

International Nuclear Information System (INIS)

Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

2010-01-01

Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

International Nuclear Information System (INIS)

Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

2003-01-01

The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Ten years's reactor operation at the Technical University Zittau - operation report

International Nuclear Information System (INIS)

Konschak, K.

1990-01-01

The Zittau Training and Research Reactor ZLFR is in use for purposes of teaching the engineers who will operate the nuclear power plants in the GDR since 10 years. Since commissioning it was started up more than 1600 times, approximately two thirds of the start-ups being utilized for purposes of teaching. A number of teaching experiments were installed that demonstrate fundamental technological processes in nuclear reactors in a manner easy to understand. The high level of nuclear safety manifests itself, among other things, in extremely low radiation exposures of the operating personal and the persons to be trained. (author)

RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

International Nuclear Information System (INIS)

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

1986-01-01

In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

RA Reactor operation and maintenance (I-IX), Part I

International Nuclear Information System (INIS)

Zecevic, V.

1963-12-01

The report on RA reactor operation and maintenance for year 1963 is divided in six tasks. This volume contains the introductory report, and three tasks of the final report, namely reactor exploitation, reactivity changes of the RA reactor before repair, planning of refuelling

RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

International Nuclear Information System (INIS)

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

1982-12-01

Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

Application of noise analysis on the operation of the RSG-GAS reactor

International Nuclear Information System (INIS)

Surian P, Tukiran S

1999-01-01

The RSG-GAS reactor has been operating for radioisotope production and experiments so that it is necessary to perform analysis of the reactor operation. analysis was done based on reactor noise experiment. Neutron noise at low and high power of the RSG-GAS has been analyzed using time and frequency domain with aim to determine the safety of reactor operation. The safety of reactor operation based on two parameters as, prompt neutron decay constant and decay ratio. The parameters are useful for reactor operation, so it is necessary to determine accurately. For determining prompt neutron decay constant, neutron density in the core of reactor which operated at 10 k W, was collected by using Fission Chamber detectors (FC). Based on power spectral density (PSD) was achieved break frequency about 23 Hz, so that the prompt neutron decay constant is about 151 sec -1 . While at at high power 20 MW, neutron density was collected by using Compensated Ionization Chamber (CIC) detector. The result at high power showed that there is reactivity effect in the core because of fluctuation in temperature and density of the coolant, and the decay ratio of 0.20, showed that the reactor is still operation in stable

Risk-assessment techniques and the reactor licensing process

International Nuclear Information System (INIS)

Levine, S.

1979-01-01

A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

Regulatory Risk Reduction for Advanced Reactor Technologies - FY2016 Status and Work Plan Summary

International Nuclear Information System (INIS)

Moe, Wayne Leland

2016-01-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy's (DOE) Advanced Reactor Technologies (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants

Method of reactor operation

International Nuclear Information System (INIS)

Nakajima, Takeshi

1988-01-01

Purpose: To minimize the power change due to the increase in xenone and power distribution after reaching the rated power in the case of using fresh fuels no requiring conditioning operation thereby starting the nuclear reactor in a short period of time and stably. Method: When control rods are entirely inserted only with a purpose for the compensation of the reactivity in a xenon-unsaturated state such as upon starting of the nuclear reactor, peaking is generated in the lower portion of the reactor core. Therefore, it is necessary to insert control rods for additionally suppressing the peaking in the lower portion of the reactor core to a relatively shallow level. In view of the above, a plurality of control rods are divided into a first control rod group finally inserted in the rated power state and a second control rod group other than the above. Then, the power is once elevated to the rated power level by means of such an intermediate control rod pattern that the ratio of the total extraction amount between the first control rod group and the second control rod group is made constant. Then, the control rods are extracted stepwise while setting the ratio of the total extraction amount constant in accordance with the change of the accumulating amount of xenone, to thereby obtain the purpose. (kamimura, M.)

Good practices in heavy water reactor operation

International Nuclear Information System (INIS)

2010-06-01

The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Estimation of the radiation risks for population in the process of the BN-600 power unit operation at the Beloyarsk NPP

International Nuclear Information System (INIS)

Koltik, I.I.; Oshkanov, N.N.

2005-01-01

Dose burdens on the population are the main criterion in estimating the radiological risks during NPP operation. Results of analysis of annual dose burdens on the population in the period of the BN-600 unit operation are presented. Data on individual and collective doses due to gas-aerosol and liquid effluents of radionuclides from the BN-600 unit on critical groups of population are presented. Data on collective doses due to other types of reactors are provided. It is shown that the risks stemming from the BN type reactors are approximately 2 orderers below the risks of channel-type and WWER-reactors [ru

Radiological monitoring related to the operation of PUSPATI's Triga Reactor

International Nuclear Information System (INIS)

Fatimah Mohamad Amin; Mohamad Yusof Mohamad Ali; Lau How Mooi; Idris Besar.

1983-01-01

Reactor operation is one of the main activities carried out at the Tun Ismail Atomic Research Centre (PUSPATI) which requires radiological monitoring. This paper describes the programme for radiological monitoring which is related to the operation of the 1 MW Triga MK II research reactor which was commissioned in July, 1982. This programme includes monitoring of the radiation and contamination levels of the reactor and its associated facilities and environmental monitoring of PUSPATI's site and its environs. The data presented in this paper covers the period between 1982 to 1983 which includes both the pre-operational and operational phases of the monitoring programme. (author)

Operational power reactor health physics

International Nuclear Information System (INIS)

Watson, B.A.

1987-01-01

Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

BEACON TSM application system to the operation of PWR reactors

International Nuclear Information System (INIS)

Lozano, J. A.; Mildrum, C.; Serrano, J. F.

2012-01-01

BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

78 FR 69139 - Physical Security-Design Certification and Operating Reactors

Science.gov (United States)

2013-11-18

... Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section..., ``Physical Security--Design Certification and Operating Reactors.'' The public comment period was originally....regulations.gov and search for Docket ID NRC-2013-0225. Address questions about NRC dockets to Carol Gallagher...

Building reactor operator sustain expert system with C language integrated production system

International Nuclear Information System (INIS)

Ouyang Qin; Hu Shouyin; Wang Ruipian

2002-01-01

The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

International Nuclear Information System (INIS)

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

1989-01-01

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

Development of operation management database for research reactors

International Nuclear Information System (INIS)

Zhang Xinjun; Chen Wei; Yang Jun

2005-01-01

An Operation Database for Pulsed Reactor has been developed on the platform for Microsoft visual C++ 6.0. This database includes four function modules, fuel elements management, incident management, experiment management and file management. It is essential for reactor security and information management. (authors)

Operational transparency: an advanced safeguards strategy for future on-load refuelled reactors

International Nuclear Information System (INIS)

Whitlock, J.J.; Trask, D.

2012-01-01

The IAEA's system for tracking fuel movement in an on-load refuelled heavy-water reactor is robust, but an opportunity remains to exploit the wealth of data streaming from the reactor vault during operation and provide real-time, third-party monitoring of reactor status and history. This concept of Operational Transparency would require that large amounts of operational data be reduced in near-real time to a small subset of high-level information. Operational Transparency would enhance the IAEA's ability to monitor the state of the core to an unprecedented level. This paper provides an overview of the novel concept of Operational Transparency in heavy water reactors, using potential application to CANDU reactors as an example, and explores some of the technical challenges that will need to be solved for efficient implementation. (author)

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Raickovic, N.; Radivojevic, J.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1983-12-01

After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW [sr

TRIGA reactor operating experience

International Nuclear Information System (INIS)

Anderson, T.V.

1970-01-01

The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

A novel approach to the design and operation scheduling of heterogeneous catalytic reactors

International Nuclear Information System (INIS)

Ghodasara, Kamlesh; Smith, Robin; Hwang, Sungwon

2014-01-01

A number of studies have been conducted to reduce the overall level of catalyst deactivation in heterogeneous catalytic reactors, and improve the performance of reactors, such as yield, conversion or selectivity. The methodology generally includes optimization of the following: (1) operating conditions of the reaction system, such as feed temperature, normal operating temperature, pressure, and composition of feed streams; (2) reactor design parameters, such as dimension of the reactor, side stream distribution along the axis of the reactor beds, the mixing ratio of inert catalyst at each bed; and (3) catalyst design parameters, such as the pore size distribution across the pellet, active material distribution, size and shape of the catalyst, etc. Few studies have examined optimization of the overall catalyst reactor performance throughout the catalyst lifetime, considering catalyst deactivation. Furthermore, little attention has been given to the impact of various configurations of reactor networks and scheduling of the reactor operation (i.e., online and offline-regeneration) on the overall reactor performance throughout the catalyst lifetime. Therefore, we developed a range of feasible sequences of reactors and scheduling of reactors for operation and regeneration, and compared the overall reactor performance of multiple cases. Furthermore, a superstructure of reactor networks was developed and optimized to determine the optimum reactor network that shows the maximum overall reactor performance. The operating schedule of each reactor in the network was considered further. Lastly, the methodology was illustrated using a case study of the MTO (methanol to olefin) process

Safe operation and maintenance of research reactor

International Nuclear Information System (INIS)

Munsorn, S.

1999-01-01

The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

Safe operation and maintenance of research reactor

Energy Technology Data Exchange (ETDEWEB)

Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

1999-10-01

The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

Study on operational aspect of natural circulation HLMC reactor (1)

International Nuclear Information System (INIS)

Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

2000-08-01

The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

Improvement of pulsing operation performance in the Nuclear Safety Research Reactor (NSRR)

International Nuclear Information System (INIS)

Kobayasi, S.; Ishijima, K.; Tanzawa, S.; Fujishiro, T.; Horiki, O.

1990-01-01

The Nuclear Safety Research Reactor (NSRR) is one of the TRIGA-type research reactors widely used in the world, and has mainly been used for studying reactor fuel behaviour during postulated reactivity-initiated accidents (RIAs). Its limited pulsing operation capability, however, could produce only a power burst from low power level simulating an RIA event from essentially zero power level. A computerized automatic reactor control system was developed and installed in the NSRR to simulate a wide range of abnormal events in nuclear power plants. This digitalized reactor control system requires no manipulation of the control rods by reactor operators during the course of the pulsing operation. Using this fully automated operation system, a variety of power transients such as power ramping, power bursts from high power level, and so on were made possible with excellent stability and safety. The present modification work in the NSRR and its fruitful results indicate new possibilities in the utilization of the TRIGA type research reactor

Limits and conditions for permanent operation of the LVR-15 reactor

International Nuclear Information System (INIS)

1995-03-01

The ''Limits and Conditions'' have been set up to demonstrate compliance with all nuclear safety requirements specified by the Decree of the Czechoslovak Atomic Energy Commission No. 9/1985, and to serve as a guide for the reactor operators to respond adequately to deviations from the required values in normal operation conditions. The publications is divided into the following chapters: (1) Safety limits of the LVR-15 reactor; (2) Adjusting the reactor protection and control system parameters; (3) Limiting conditions for normal operation modes; (4) Control requirements; and (5) reactor operation management. A document by the Institute for Research, Production, and Application of Radioisotopes specifying conditions for the acceptance of radioactive wastes to the central repository is appended. (P.A.). 7 tabs

Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

International Nuclear Information System (INIS)

1986-11-01

This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

Proceedings of the international symposium on research reactor safety operations and modifications

International Nuclear Information System (INIS)

1990-03-01

The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

Can a nuclear reactor operate for 100 years?

International Nuclear Information System (INIS)

Hertel, O.

2010-01-01

The TWR (Travelling Wave Reactor) concept was invented in the fifties, then forgotten and it reappeared in 2001 but it was considered too immature to be selected for the fourth generation of nuclear reactors, now an American company 'Terrapower' proposes one whose design is given in the article. This TWR operates with depleted uranium, only the lower part of the fuel rod involves uranium fuel with a civil enrichment ratio (less that 20%). The lower part of the fuel will ignite the fission reaction and enrich the part of fuel just above through neutron absorption. The burning part of the fuel will move up progressively. The main advantage of this reactor is that it can operate for decades without maintenance nor fuel loading. The principle is right on the paper but requires huge technological work to select materials and systems that will be able to withstand decades of operation time in harsh conditions. (A.C.)

Comparative Study on Research Reactor Operation Requirements of IAEA and Korea

International Nuclear Information System (INIS)

Yune, Young Gill; Chang, Woo Joon; Lee, Jinho

2014-01-01

The IAEA safety requirements represent a consensus view of the IAEA's member states and encompass the recommendations of international experts for the safety of the research reactor. The IAEA safety standards are acknowledged as a global standard by most countries in the world. Therefore, it is desirable to harmonize domestic safety requirements with those of the IAEA. Also, since the IAEA safety requirements include international endeavors to strengthen nuclear safety and to apply the latest nuclear safety technologies to nuclear facilities, a comparative analysis of the safety requirements of the IAEA and Korea for the research reactor will be beneficial to obtaining the insights to improve regulations of the research reactor in Korea. For the reason, a comparative study has been conducted for the research reactor operation safety requirements of the IAEA and Korea in this paper. This paper briefly introduced operation safety requirements in the NS-R-4, analysis methods of this study, and the analysis results. A comparative study has been performed on research reactor operation safety requirements of the IAEA and Korea. This study has identified some gaps in operation safety requirements for the NPP of Korea, comparing with those for the research reactor of the IAEA in the following areas: periodic safety assessment, utilization and modification of the reactor, extended shutdown, inspection/testing/maintenance, operating organization, emergency planning, and commissioning

Operating history report for the Peach Bottom HTGR. Volume I. Reactor operating history

International Nuclear Information System (INIS)

Scheffel, W.J.; Baldwin, N.L.; Tomlin, R.W.

1976-01-01

The operating history for the Peach Bottom-1 Reactor is presented for the years 1966 through 1975. Information concerning general chemistry data, general physics data, location of sensing elements in the primary helium circuit, and postirradiation examination and testing of reactor components is presented

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

International Nuclear Information System (INIS)

1978-09-01

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

Energy Technology Data Exchange (ETDEWEB)

Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-08-01

Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologies (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants

Risk assessment of a fusion-reactor fuel-processing system

International Nuclear Information System (INIS)

Bruske, S.Z.; Holland, D.F.

1983-07-01

The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

2011-07-01

The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

Risk/benefits of fast breeder reactors societal risks and accident modelling

International Nuclear Information System (INIS)

Farmer, F.R.

1979-01-01

A special feature of the Fast Breeder Reactor is the possibility of fuel vaporisation, hence accidents may have more severe consequences than thermal reactors. This article discusses the process of accident modelling, the identification and assessment of risk, not yet incurred. 10 refs

ASSESSMENT OF THE POTENTIAL FOR HYDROGEN GENERATION DURING GROUTING OPERATIONS IN THE R AND P REACTOR VESSELS

Energy Technology Data Exchange (ETDEWEB)

Wiersma, B.

2010-05-24

The R- and P-reactor buildings were retired from service and are now being prepared for deactivation and decommissioning (D and D). D and D activities consist primarily of immobilizing contaminated components and structures in a grout-like formulation. Aluminum corrodes very rapidly when it comes in contact with the alkaline grout materials and as a result produces hydrogen gas. To address this potential deflagration/explosion hazard, the Materials Science and Technology Directorate (MS and T) of the Savannah River National Laboratory (SRNL) has been requested to review and evaluate existing experimental and analytical studies of this issue to determine if any process constraints on the chemistry of the fill material and the fill operation are necessary. Various options exist for the type of grout material that may be used for D and D of the reactor vessels. The grout formulation options include ceramicrete (pH 6-8), low pH portland cement + silica fume grout (pH 10.4), or Portland cement groupt (pH 12.5). The assessment concluded that either ceramicrete or the silica fume grout may be used to safely grout the P-reactor vessel. The risk of accumulation of a flammable mixture of hydrogen between the grout-air interface and the top of the reactor is very low. Portland cement grout, on the other hand, for the same range of process parameters does not provide a margin of safety against the accumulation of flammable gas in the reactor vessel during grouting operations in the P-reactor vessel. It is recommended that this grout not be utilized for this task. The R-reactor vessel cotnains significantly less aluminum based on current facility process knowledge, surface observations, and drawings. Therefore, a Portland cement grout may be considered for grouting operations as well as the other grout formulations. For example, if the grout fill rate is less than 1 inch/min and the grout temperature is maintained at 70 C or less, the risk of hydrogen accumulation during fill

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

Energy Technology Data Exchange (ETDEWEB)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu [Harbin Engineering University, Harbin (China)

2014-08-15

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions.

The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

International Nuclear Information System (INIS)

Fang, Maoyao; Peng, Minjun; Cheng Shouyu

2014-01-01

In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions

A logic scheme for regulating safe operation of research reactors

Energy Technology Data Exchange (ETDEWEB)

Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Effat, A; Rahman, F A [Operational Saety Dept, National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

This investigation presents a logic scheme for regulating the safe operation of research reactor in accordance with the new revision of SS-35 and revised by the 10 CFR. It emphasizes the regulatory inspection and enforcement (RI end E) during the reactor operation phase. IT is developed to provide information, guidance and recommendations to be taken when constructing the RI and E program that could be applied to the operational phase of the egyptian Research Reactors. In the operational phase, the regulatory inspection (RI) means an examination, observation, measurement, or test undertaken or on behalf of the nuclear regulatory body (NRB) during operation to verify that the nuclear materials, components, systems and structures as well operational activities, processes, procedures and personnel competence and performance are in accordance with the requirements established or the provisions approved by NRB or specified in the operational license or contained in regulations. Regulatory inspection includes both routine and non-routine ones. Any of them may be announced or unannounced. The problems identified by the RI must be resolved by the proper RE actions. The RE actions include investigative and corrective RE actions. These RI and E procedures for regulating safe operation of research reactors are presented as flow charts and then developed as a computer logic scheme. The software program is very efficient, very friendly, very simple and is interactive in nature such that the program asks the user certain questions about essential steps that guide the (RI and E) for research reactors, and user responds. The program proceeds based on this response until all the necessary steps for (RI and E) are accomplished. 5 figs.

Operational limits and conditions and operating procedures for research reactors. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This publication provides practical guidance on all important aspects of developing, formulating and presenting the operational limits and conditions as well as the operating procedures for research reactors. It covers the concept of operational limits and conditions, their content, and the responsibilities of the operating organization with respect to their establishment, modification, documentation and compliance. The guidance also covers the training of operating personnel on performing periodic testing, established by the operational limits and conditions, and operating procedures

Current status of operation and utilization of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Le Van So

2004-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor using the Soviet WWR-SM fuel assembly with 36% enrichment of U-235. It was upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analyses and research purposes. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for physics experiments and training purpose. From the first start-up to the end of December 2002, it totaled about 24,700 hrs of operation and the total energy released was 490 MWd. After 10 years of operation with the core of 89-fuel assembly configuration, in April 1994, the first refueling work was done and the 100-fuel assembly configuration was set-up. The second fuel reloading was executed in March 2002. At present time, the working configuration of the reactor core consists of 104 fuel assemblies. This fuel reloading will ensure efficient exploitation of the reactor for about 3 years with 1200-1300 hrs per year at nominal power. The current status of operation and utilization and some activities related to the reactor core management of the DNRR are presented and discussed in this paper. (author)

Some post operational adjustments to the prototype fast reactor at Dounreay

International Nuclear Information System (INIS)

Lunt, A.R.W.

1979-01-01

Prior to and during the initial operation of the Prototype Fast Reactor at Dounreay certain features have been considered to be in need of adjustment to provide better operating characteristics. This article describes the work done to support the consequential changes of operational techniques and plant design in the following areas: maintenance of dry conditions at the superheater steam inlets, the temperature control of the reactor roof, and the introduction of a system enabling the reactor to continue running after a turbine trip. (author)

Design requirements, operation and maintenance of gas-cooled reactors

International Nuclear Information System (INIS)

1989-06-01

At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

Current status of operation and utilization of the Dalat Research Reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi

2006-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the HEU (36% enrichment) WWR-M2 fuel assemblies. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in the 1st November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2004, it totaled about 27,253 hrs of operation and the total energy released was about 543 MWd. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 fuel assemblies (FA). The 11 new FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 FAs. The second fuel reloading was executed in March 2002. The 4 new FAs were added in the core periphery, at previous beryllium element locations. The working configuration of 104 FAs ensured efficient exploitation of the DNRR at nominal power for about 3000 hrs since March 2002. In order to provide excess reactivity for the reactor operation without the need to discharge high burned FAs, in June 2004, the fuel shuffling of the reactor core was done. 16 FAs with low burn-up from the core periphery were moved toward the core center and 16 FAs with high-burn-up from the core center were moved toward the core periphery. This operation provided additional reactivity of about 0.85 β eff that the current reactor configuration using re-shuffled HEU fuel is expected to allow normal operation until June 2006. In 1999, the request of returning to Russia HEU fuels from foreign

Reactor siting risk comparisons related to recommendations of NUREG-0625

International Nuclear Information System (INIS)

Barsell, A.W.; Dombek, F.S.; Orvis, D.D.

1980-11-01

This document evaluates how implementing the remote siting recommendations for nuclear reactors (NUREG-0625) made by the Siting Policy Task Force of the US Nuclear Regulatory Commission (NRC) can reduce potential public risk. The document analyzes how population density affects site-specific risk for both light water reactors (LWRs) and high-temperature gas-cooled reactors

TARMS, an on-line boiling water reactor operation management system

International Nuclear Information System (INIS)

Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

1984-01-01

The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de

Experimental evaluation of an expert system for nuclear reactor operators

International Nuclear Information System (INIS)

Nelson, W.R.

1984-10-01

The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

The Influence of RSG-GAS Primary Pump Operation Concerning the Rise Water Level of Reactor Pool in 15 MW Reactor Power

International Nuclear Information System (INIS)

Djunaidi

2004-01-01

The expansion of air volume in the delay chamber shows in rise water level of reactor pool during the operation. The rises of water level in the reactor pool is not quite from the expansion of air volume in the delay chamber, but some influence the primary pump operation. The purpose evaluated of influence primary pump is to know the influence primary pump power concerning the rise water level during the reactor operation. From the data collection during 15 MW power operation in the last core 42 the influence of primary pump operation concerning the rise water level in the reactor pool is 34.48 % from the total increased after operation during 12 days. (author)

Safety margins of operating reactors. Analysis of uncertainties and implications for decision making

International Nuclear Information System (INIS)

2003-01-01

Maintaining safety in the design and operation of nuclear power plants (NPPs) is a very important task under the conditions of a challenging environment, affected by the deregulated electricity market and implementation of risk informed regulations. In Member States, advanced computer codes are widely used as safety analysis tools in the framework of licensing of new NPP projects, safety upgrading programmes of existing NPPs, periodic safety reviews, renewal of operating licences, use of the safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, for justification of lifetime extensions, development of new emergency operating procedures, analysis of operational events, and development of accident management programmes. The issue of inadequate quality of safety analysis is becoming important due to a general tendency to use advanced tools for better establishment and utilization of safety margins, while the existence of such margins assure that NPPs operate safely in all modes of operation and at all times. The most important safety margins relate to physical barriers against release of radioactive material, such as fuel matrix and fuel cladding, reactor coolant system boundary, and the containment. Typically, safety margins are determined with use of computational tools for safety analysis. Advanced best estimate computer codes are suggested e.g. in the IAEA Safety Guide on Safety Assessment and Verification for Nuclear Power Plants to be used for current safety analysis. Such computer codes require their careful application to avoid unjustified reduction in robustness of the reactor safety. The issue of uncertainties in safety analyses and their impact on evaluation of safety margins is addressed in a number of IAEA guidance documents, in particular in the Safety Report on Accident Analysis for Nuclear Power Plants. It is also discussed in various technical meetings and workshops devoted to this area. The

Reactor operations for nuclear pumping of lasers

Energy Technology Data Exchange (ETDEWEB)

Beck, G; Cooper, G [University of Illinois (United States)

1974-07-01

Experiments involving the measurement of gas parameters that are related to lasing, and lasing of various gas mixtures have comprised a major part of the utilization of the University of Illinois Advanced TRIGA Reactor since the upgrading of the facility was completed in 1969. A thru beam port, which was added during upgrading, has been the facility used for these measurements. The laser cell is placed in the port adjacent to the core. Alignment is then accomplished by using both ends of the port or by a mirror placed at the back side of the apparatus. The reactor has been operated in all modes (pulsing, square wave, and steady state) for the experiments although pulsing is the primary mode that is used. Laser enhancement has been obtained in several cases, but efforts toward direct pumping from the radiation alone have not as yet succeeded. Improved laser operation from direct pumping has been suggested with an emphasis on high-powered systems where the basic input energy is to be derived from a nuclear reactor.

Operational and passive safety aspects of the STAR-LM natural convection HLMC reactor. Study on operational aspects of a natural circulation HLMC reactor. 2

International Nuclear Information System (INIS)

Sienicki, J.J.; Petkov, P.V.

2001-09-01

The concept of a heavy liquid metal cooled fast reactor that achieves 100+% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100+% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the earlier Phase 1 of the project, the stage for the overall study was prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code was developed that has the capabilities to calculate operational and accident transients. Code input was prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change-in-turbine load demonstrated the capability to analyze typical transient cases. The present second phase of the study involves documentation of the plant dynamics analysis computer code including major assumptions and thermal hydraulic equations as well as application of the code to calculate operational transients and postulated accidents. The following normal and accident scenarios are calculated: initial startup; normal shutdown; startup from hot standby; decrease-in-turbine load; increase-in-turbine load; loss-of-heat sink without scram; overcooling event without scram; and unprotected transient overpower. For the decrease

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

A series of lectures on operational physics of power reactors

International Nuclear Information System (INIS)

Mohanakrishnan, P.; Rastogi, B.P.

1982-01-01

This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Graphics and control for in-reactor operations

International Nuclear Information System (INIS)

Smith, A.L.

1996-01-01

Remotely operated inspection, repair and maintenance tasks for Magnox Electric's eight twin MAGNOX-type reactor stations must conform to the Nuclear Installations Inspectorate's regulations. The performance of manipulator systems used to undertake such operations has improved significantly with the recent introduction of computer graphics. These are used for simulation and demonstration purposes as well as a real-time aid to the manipulator operator. Outage times can be reduced by the appropriate use of such computer technology to reduce manipulator operation times. (UK)

Supervisory and managerial aspects on the safe operation of the Egyptian second research reactor

International Nuclear Information System (INIS)

Abdelrazek, I.D.; Shokr, A.M.

2000-01-01

ETRR-2 is a multipurpose reactor for radioisotopes production, neutron beam experiments, basic and applied research in physics and engineering, fuels and material tests, and for training. The reactor is an open pool type, 22 MW power, with average thermal neutron flux of 1.4x10 14 n/cm 2 sec, cooled and moderated by light water and with beryllium reflectors. Various experimental devices and irradiation facilities are integrated with the reactor. The reactor has been licensed for operation in November 1998. Several principles and regulations have been applied to all the reactor project stages to achieve safety. Moreover, other several principals, regulations, and aspects are enforced by the AEA, National Center for Nuclear Safety and Radiation Control, NCNSRC, and reactor management to achieve safety during reactor operation and utilization. Responsibility on Safety and Supervision Aspects AEA chairman has the ultimate responsibility on the reactor safety during operation and utilization. The primary responsible on the safety is the ETRR-2 supervisor, who is supervising the ETRR-2 site that includes the ETRR-2 reactor, Fuel Manufacturing Plant, and other two projects (under execution): Radioisotopes Production Plant and Dry Fuel Storage Facility. ETRR-2 supervisor is responsible to ensure that: the reactor is operated in accordance with the safety requirements by qualified and trained personnel, updating and enforcement of the reactor mandatory documentation, and the services are adequate for the reactor operation. He is responsible, also, to guarantee that: the reactor manager has enough authority and resources to carry out his function effectively and the reactor is kept operated in agreement with established procedures. A Technical Revision Committee, TRC, is formed to advice the ETRR-2 supervisor on the safety of the reactor and experiments. The committee members are from AEA experts with no direct relation to the reactor and experiments being performed. Members

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.

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

Operation of the SLOWPOKE-2 reactor in Jamaica

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

Current status of operation and utilization of the Dalat Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien

2006-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the Soviet VVR-M2 fuel assembly with 36% enrichment of U-235. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2003, it totaled about 26,000 hrs of operation and the total energy released was about 515 MWd. After 10 years of operation with the core of 89-fuel assembly configuration, in April 1994, the first refueling work was done and the 100-fuel assembly configuration was set-up. The second fuel reloading was executed in March 2002. At present time, the working configuration of the reactor core consists of 104 fuel assemblies. The next fuel reloading has been planned at the end of 2004. The current status of operation and utilization of the DNRR is presented and discussed in this paper. (author)

RA reactor operation and maintenance, Annual report 1974

International Nuclear Information System (INIS)

Milosevic, D. et al

1974-12-01

During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m 2 of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems [sr

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. A phased approach was used in Level 1. In Phase 1 the concept of plant operational states (POSs) was developed to provide a better representation of the plant as it transitions from power to non power operation. This included a coarse screening analysis of all POSs to identify vulnerable plant configurations, to characterize (on a high, medium, or low basis) potential frequencies of core damage accidents, and to provide a foundation for a detailed Phase 2 analysis. In Phase 2, selected POSs from both Grand Gulf and Surry were chosen for detailed analysis. For Grand Gulf, POS 5 (approximately Cold Shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected. For Surry, three POSs representing the time the plant spends in mid loop operation were chosen for analysis. Level 1 and Level 2/3 results from the Surry analyses are presented

Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

Energy Technology Data Exchange (ETDEWEB)

Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

2006-07-01

A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976

International Nuclear Information System (INIS)

Martinc, R.

1976-12-01

During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation [sr

RA Research reactor, Annual report 1970 - Operation and maintenance

International Nuclear Information System (INIS)

Milosevic, D. et al.

1970-12-01

During 1970, the RA Reactor was operated at nominal power of 6.5 MW for 160 days, and 40 days at lower power levels. Total production mounted to 25968 MWh which is 3.87% higher than planned. The action plan was changed compared to the previous years because of sending the heavy water to France for re-concentration. Isotopic concentration of the heavy water was decreased to 99.05% and now after re-concentration it is 99.96%. Discrepancy from the action plan, in September was caused by the delay return of the heavy water for administrative and transportation difficulties. The restart of the reactor in September was postponed because the cladding of one fuel element was damaged immediately after the start-up, and the reactor had to be shutdown. In October and November reactor was in operation 28 and 25 days respectively which enabled to make up for the lost time. Reactor was used for irradiation and experiments according to the demand of 390 users, 340 from the Institute and 50 external users. This report contains detailed data about reactor power and experiments performed in 1969. It is concluded that the reactor operated successfully according to the plan. Shorter interruptions were caused only by difficulties with water supply pipes and sliding of the soil. Reactor was only twice scram shutdown because of the false signals caused by failures of the electronic control instrumentation. the period when reactor was not in operation was used for inspection of the reactor vessel internals. By using special TV cameras and telescopes, it was found that the there are no signs of corrosion on the reactor vessel, e.e. that the internals are in a very good state. Simultaneously, connection for the pipes of future emergency core cooling system were constructed. During 1970, the spent fuel was repacked from fuel channels into special aluminium casks. Four casks containing 660 fuel slugs was deposited int the storage pool No.4. There is now 18 casks with 2951 spent fuel slugs in

Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

International Nuclear Information System (INIS)

Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

1987-01-01

The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

Core design of long life-cycle fast reactors operating without reactivity margin

International Nuclear Information System (INIS)

Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

2012-01-01

In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

Constrained model predictive control for load-following operation of APR reactors

International Nuclear Information System (INIS)

Kim, Jae Hwan; Lee, Sim Won; Kim, Ju Hyun; Na, Man Gyun; Yu, Keuk Jong; Kim, Han Gon

2012-01-01

The load-following operation of APR+ reactor is needed to control the power effectively using the control rods and to restrain the reactivity control from using the boric acid for flexibility of plant operation. Usually, the reason why the disproportion of axial flux distribution occurs during load-following operation is xenon-induced oscillation. The xenon has a very high absorption cross-section and makes the impact on the reactor delayed by the iodine precursor. The power maneuvering using automatically load-following operation has advantage in terms of safety and economic operation of the reactor, so the controller has to be designed efficiently. Therefore, an advanced control method that meets the conditions such as automatic control, flexibility, safety, and convenience is necessary to load-following operation of APR+ reactor. In this paper, the constrained model predictive control (MPC) method is applied to design APR reactor's automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control. Some controllers use only the current tracking command, but MPC considers future commands in addition to the current tracking command. So, MPC can achieve better tracking performance than others. Furthermore, an MPC is to used in many industrial process control systems. The basic concept of the MPC is to solve an optimization problem for a finite future time interval at present time and to implement the first optimal control input as the current control input. The KISPAC-1D code, which models the APR+ nuclear power plants, is interfaced to the proposed controller to verify the tracking performance of the reactor power level and ASI. It is known that the proposed controller exhibits very fast tracking responses

Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

International Nuclear Information System (INIS)

Moreira, Joao Manoel Losada; Frajndlich, Roberto

1995-01-01

This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

International Nuclear Information System (INIS)

Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

1986-09-01

Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m 2 and a surface heat flux of 1 MW/m 2 . The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO 2 rods. The helium coolant pressure is 5 MPa, entering the module at 297 0 C and exiting at 550 0 C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

Indian Academy of Sciences (India)

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

N reactor external events probabilistic risk assessment

International Nuclear Information System (INIS)

Baxter, J.T.

1989-01-01

An external events probabilistic risk assessment of the N Reactor has been completed. The methods used are those currently being proposed for external events analysis in NUREG-1150. Results are presented for the external hazards that survived preliminary screening. They are earthquake, fire, and external flood. Core damage frequencies for these hazards are shown to be comparable to those for commercial pressurized water reactors. Dominant fire sequences are described and related to 10 CFR 50, Appendix R design requirements. Potential remedial measures that reduce fire core damage risk are described including modifications to fire protection systems, procedure changes, and addition of new administrative controls. Dominant seismic sequences are described. The effect of non-safety support system dependencies on seismic risk is presented

Simulator platform for fast reactor operation and safety technology demonstration

International Nuclear Information System (INIS)

Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

2012-01-01

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Simulator platform for fast reactor operation and safety technology demonstration

Energy Technology Data Exchange (ETDEWEB)

Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

International Nuclear Information System (INIS)

1995-08-01

This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory

Clustering methods and visualization algorithms to aid nuclear reactor operative diagnostics

International Nuclear Information System (INIS)

Pepelyshev, Yu.N.; Dzwinel, W.

1990-01-01

The software system developed plays the role of the aid to an operator for nuclear reactor diagnostics. The noise analysis of the reactor parameters such as power, temperature and coolant flow rate constitutes the basis of the system. Combination of data acquisition, data preprocessing, clustering and cluster visualization algorithms with heuristic techniques of results analysis, determine the way of its implementation. Two regimes are available. The first one - extended - is recommended for a long term investigations and the second - suppressed for the aid to the reactor operation monitoring. The system has been tested and developed at the JINR IBR-2 pulsed reactor. 13 refs.; 4 figs.; 2 tabs

Steady-state operation requirements of tokamak fusion reactor concepts

International Nuclear Information System (INIS)

Knobloch, A.F.

1991-06-01

In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

Cladding failure probability modeling for risk evaluations of fast reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Kramer, J.M.

1987-01-01

This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

Operation and maintenance of the RB reactor, Annual report for 1980

International Nuclear Information System (INIS)

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

1980-12-01

This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 9 Annexes as follows: Utilization of the RB reactor from 1976 - 1980; program of reactor utilization from 1981-1985; contents of the RB reactor safety report; maintenance of the reactor components and equipment in 1980; verification of reactor reliability after the earthquake (May 18 1980); refurbishment of equipment in 1980, and purchasing new equipment from 1981-1985; review of radiation doses in the reactor building and exposure doses for the reactor staff; personnel data and financial data

Artificial Intelligence and its Reasonable Application Scenario to Reactor Operation

International Nuclear Information System (INIS)

Im, Ki Hong; Suh, Yong-Suk; Park, Cheol; Lim, In-Cheol

2017-01-01

This paper presents brief but reasonable scenarios for applying AI or machine learning technologies to research reactor from various perspectives. Two less safety critical scenarios for applying AI to reactor operation are introduced in this study. However, the AI assistant will not only be an assistant but it will also be an operator in the future. What is required is big operation data which can represent all the cases requiring operation decision, including normal operation and accident data as well, and enough time to train and fix the AI system with this data. We can predict AI study in this area can begin with a mild and safe application. But in the near future, this technology could be used to handle or automate more severe operations.

Reverse flow operation with reactor side feeding : analysis, modeling and simulation

NARCIS (Netherlands)

Budhi, Y.W.; Hoebink, J.H.B.J.; Schouten, J.C.

2004-01-01

The novel concept of reverse flow operation with reactor side feeding is studied for selective oxidation of NH3 to produce either N2, N2O, or NO. During normal reverse flow operation, where the feeds are alternately introduced from either end of the reactor, the conversion is always lower when

Ignition and time-dependent fractional power operation of tokamak reactors

International Nuclear Information System (INIS)

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-01-01

The eventual utilization of a tokamak fusion reactor for commercial power necessitates a thorough understanding of the operational requirements at full and fractional power levels and during transitions from one operating level to another. In this study we examine the role of burn control in maintaining the reactor plasma at equilibrium to avoid thermal runaway during fractional power operation. Because these requirements rely so heavily on the assumptions that govern the plasma transport, this study focuses on time-dependent analyses and a comparison of ignition requirements using a range of energy confinement

Annual report on operation, utilization and technical development of research reactors and hot laboratory

International Nuclear Information System (INIS)

1990-09-01

This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

The heating operational summarization in three winters of a 5 MW test heating reactor

International Nuclear Information System (INIS)

Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

1992-09-01

The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

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

International Nuclear Information System (INIS)

Bintu Khoiriyyah; Budi Purnama; Tri Cahyo Laksono

2016-01-01

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

Advantages of forced non-steady operated trickle-bed reactors

NARCIS (Netherlands)

Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

2002-01-01

Trickle-bed reactors are usually operated in the steady state trickle flow regime. Uneven liquid distribution and the formation of hot spots are the most serious problems experienced during trickle flow operation. In this paper, we advocate the use of non-steady state operation of trickle-bed

Operating reactors licensing actions summary. Volume 5, No. 9

International Nuclear Information System (INIS)

1985-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Volume 5, No. 8

International Nuclear Information System (INIS)

1985-10-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Volume 4, No. 9

International Nuclear Information System (INIS)

1984-11-01

This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Operating reactors licensing actions summary. Vol. 4, No. 2

International Nuclear Information System (INIS)

1984-04-01

This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR)

International Nuclear Information System (INIS)

Gilmore, W.E.; Blackman, H.S.; Ryan, T.G.

1986-01-01

The Nuclear Computerized Library for Assessing Reactor Risk (NUCLARR) program is a multiyear effort sponsored by the NRC and is being conducted at the Idaho National Engineering Laboratory (INEL). The goal of this program is to establish and operate computerized data base management tools for the human reliability data bank specification developed by Comer and Donovan. The NRC and the risk analysis community recognized that implementing a fully functional library would not be feasible, or practical, without the aid of computerized tools for management and manipulation of its data sources. The end users of the NUCLARR can be classified into three categories according to specific needs. The first category is those users interested in reviewing individual data sources for a given situation. The second category of users selects multiple data sources for a specific case, summarizing the information, and performing comparative studies. The last category of users interfaces the NUCLARR with other programming applications, such as other data banks, and simulation models of risk assessment. Project status is provided in the paper

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

Safe Operation of Critical Assemblies and Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1961-05-15

This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

Safe Operation of Critical Assemblies and Research Reactors

International Nuclear Information System (INIS)

1961-01-01

This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

Computer-aided testing and operational aids for PARR-1 nuclear reactor

International Nuclear Information System (INIS)

Ansari, S.A.

1990-01-01

The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

International Nuclear Information System (INIS)

Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

1999-01-01

The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

Self operation type reactor scram device

International Nuclear Information System (INIS)

Saito, Makoto; Gunji, Minoru.

1992-01-01

A control rod having neutron absorbers therein is held by a curie point electromagnet by way of a control rod extension shaft. The electromagnet is suspended from a vertically movable driving shaft in an upper guide tube. Then, a heater is disposed at the lower portion in the inner side of the upper guide tube. Upon a function confirmation test, the electromagnet is at first pulled up to the inside of the upper guide tube. Subsequently, the electromagnet is heated by the heater by a temperature higher than the curie point of the temperature sensing magnetic material. If the function is normal, armature connected to the control rod extension tube is separated. With such a constitution, the electromagnetic portion is isolated from a coolant main stream, thereby enabling to avoid the cooling effect by the stream of coolants. Accordingly, the operation test for confirming the integrity of the function of the curie point electromagnet can be conducted while placing the electromagnet in the reactor core as it is during actual reactor operation. (I.N.)

Self operation type reactor control device

International Nuclear Information System (INIS)

Saito, Makoto; Gunji, Minoru.

1990-01-01

A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

The influence of the operating schedule of the Greek Research Reactor on the radiological consequences of the reactor

International Nuclear Information System (INIS)

Kollas, J.G.

1986-04-01

The sensitivity of the radiological consequences of the Greek Research Reactor to the operating schedule of the reactor is assessed in this report. The consequences are due to the occurrence of a postulated accident, a 20% core melt loss of coolant accident. Three different operating schedules are considered: (a) the present 8 hrs/day, 5 days/wk schedule, (b) a 16 hrs/day, 5 days/wk schedule, and (c) a continuous operation schedule. The results of the analysis indicate that there is a direct relation between consequences and duration of operation. (author)

Periodic Operation of Three-Phase Catalytic Reactors

Czech Academy of Sciences Publication Activity Database

Silveston, P.T.; Hanika, Jiří

2005-01-01

Roč. 82, č. 6 (2005), s. 1105-1142 ISSN 0008-4034 Institutional research plan: CEZ:AV0Z4072921 Keywords : three-phase reactors * trickle bed * periodic operation Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 0.574, year: 2005

The SM and MIR reactors operation experience

International Nuclear Information System (INIS)

Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

1995-01-01

The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

Method of operating a water-cooled nuclear reactor

International Nuclear Information System (INIS)

Lysell, G.

1975-01-01

When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

Assessment of risks of accidents and normal operation at nuclear power plants

International Nuclear Information System (INIS)

Savolainen, Ilkka; Vuori, Seppo.

1977-01-01

A probabilistic assessment model for the analysis of risks involved in the operation of nuclear power plants is described. With the computer code ARANO it is possible to estimate the health and economic consequences of reactor accidents both in probabilistic and deterministic sense. In addition the code is applicable to the calculation of individual and collective doses caused by the releases during normal operation. The estimation of release probabilities and magnitudes is not included in the model. (author)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

Energy Technology Data Exchange (ETDEWEB)

McKenzie, J L [Armed Forces Radiobiology Research Institute, Bethesda, MD (United States)

1974-07-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the {sup 41}Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Operational experience at the AFRRI-TRIGA reactor facility (1972-1974)

International Nuclear Information System (INIS)

McKenzie, J.L.

1974-01-01

The Armed Forces Radiobiology Research Institute operates a TRIGA Mark-F Reactor which has a movable core, and the capability to operate in the steady state mode up to a maximum power level of one megawatt and in the pulse mode up to a maximum peak power of 2600 MW (10 millisecond pulse). The reactor experienced three operational incidents during the period from February 1972 to February 1974, and two of these incidents were reportable to the Atomic Energy Commission. The first incident consisted of a failure of a weld at the top of the tri-flute on an instrumented fuel element which allowed the tri-flute to move up about one-half inch from its normal position. The instrumented fuel element was removed from the reactor core and replaced with a new instrumented fuel element. The second incident consisted of a malfunction of the reactor core position safety interlock which resulted in the lead shield doors closing around the reactor core shroud. The lead shield doors did not make contact with the reactor core shroud and therefore no damage occurred. The incident was reported to the Atomic Energy Commission. The third incident consisted of a failure of the threaded connector on the top of the transient control rod which allowed the transient control rod to separate from the connecting rod and drop to the bottom of the guide tube. The damaged transient control rod was removed from the guide tube and a new transient rod was installed in the reactor core. This incident was reported to the Atomic Energy Commission. A modification was made to Exposure Room 2 which consisted of placing panels, painted with gadolinium oxide paint, on the walls, ceiling, and reactor core tank projection. This resulted in the 41 Ar production rate and the effluent release to the environment being reduced by a factor of 10 to 20, depending upon the position of the reactor core. (author)

Broad-Application Test Reactor

International Nuclear Information System (INIS)

Motloch, C.G.

1992-05-01

This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

Licensed operating reactors status summary report data as of January 31, 1989

International Nuclear Information System (INIS)

Schwartz, I.

1989-03-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information, such as spent fuel storage capability, reactor years of experience, and non-power reactors in the United States

Licensed operating reactors: status summary report, data as of May 31, 1982

International Nuclear Information System (INIS)

1982-06-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 02-28-89

International Nuclear Information System (INIS)

1989-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactor Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors status summary report, data as of December 31, 1988

International Nuclear Information System (INIS)

1989-02-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report data as of October 31, 1982

International Nuclear Information System (INIS)

1982-11-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors. Status summary report, data as of 8-31-82

International Nuclear Information System (INIS)

1982-09-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report data as of 9-30-86

International Nuclear Information System (INIS)

1987-03-01

The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices. IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 11-30-88

International Nuclear Information System (INIS)

1989-01-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information, such as spent fuel storage capability, reactor years of experience, and non-power reactors in the United States

Licensed operating reactors: Status summary report, data as of 11-30-86

International Nuclear Information System (INIS)

1987-06-01

The operating units status report - licensed operating reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Administration and Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US. It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the US energy situation as a whole

Licensed operating reactors: Status summary report, Data as of 08-31-86

International Nuclear Information System (INIS)

1987-03-01

The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

Regulation concerning installation and operation of reactors for power generation

International Nuclear Information System (INIS)

1987-01-01

This report shows the Ordinance of the Ministry of International Trade and Industry No.77 of December 28, 1978. The ordinance consists of provisions covering application for permission for construction of nuclear reactor (concerning continuous maximum thermal output, location, structure, reactor core, fuel material, moderator, reflector, cooling system, measurement control system, safety circuit, control system, emergency system, radioactive waste proposal facilities, construction plan, meteorology and other environmental conditions, etc.), operation plan (to be submitted every year), application for approval of joint management (name, address, facilities location, conditions for joint management, etc.), cancellation of permission (in five years from the date of permission), record keeping (density and temperature of neutron, temperature and pressure of coolant, purity of mederator, etc.), restriction on access to areas under management (measures to be taken in such areas), measures concerning exposure to radioactive rays (allowable dosage, etc.), patrol and checking in nuclear reactor facilities, self-imposed regular inspection of nuclear reactor facilities, operation of nuclear reactor, transport within plant or business establishment, storage (storing facilities, etc.), waste disposal, etc. (Nogami, K.)

Nuclear reactor operator licensing

International Nuclear Information System (INIS)

Bursey, R.J.

1978-01-01

The Atomic Energy Act of 1954, which was amended in 1974 by the Energy Reorganization Act, established the requirement that individuals who had the responsibility of operating the reactors in nuclear power plants must be licensed. Section 107 of the act states ''the Commission shall (1) prescribe uniform conditions for licensing individuals; (2) determine the qualifications of such individuals; and (3) issue licenses to such individuals in such form as the Commission may prescribe.'' The article discusses the types of licenses, the selection and training of individuals, and the administration of the Nuclear Regulatory Commission licensing examinations

Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

International Nuclear Information System (INIS)

Rastogi, B.P.

1989-01-01

This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

Calculations for accidents in water reactors during operation at power

International Nuclear Information System (INIS)

Blanc, H.; Dutraive, P.; Fabrega, S.; Millot, J.P.

1976-07-01

The behaviour of a water reactor on an accident occurring as the reactor is normally operated at power may be calculated through the computer code detailed in this article. Reactivity accidents, loss of coolant ones and power over-running ones are reviewed. (author)

Economic evaluation of nuclear reactor operation utilizing power effect

International Nuclear Information System (INIS)

Budinsky, M.; Mydliar, J.

1988-01-01

The operation of a reactor at the so-called power effect may substantially increase the burnup of fuel to be removed. The aim of the evaluation of such reactor operation is the optimal determination of the time over which the yield of the higher use of fuel exceeds economic losses resulting from the increased share of constant expenditure of the price of generated kWh of electric power which ensues from such operation. A mathematical model is presented for such evaluation of reactor operation with regard to benefits for the national economy which is the basis of the ESTER 2 computer program. The calculations show that the prices of generated and delivered kWh are minimally 2% less than the prices of generated power without the power effect use. The minimum ranges in the interval of 30 to 50 days. The dependence of the price of generated and delivered kWh from the point of view of the operator of the power plant as well as the component of fuel price of generated kWh will not reach the minimum even after 50 days of operation. From the operating and physical points of view the duration of power effect is not expected to exceed 20 to 30 days which means that from the point of view of the national economy the price of generated and delivered kWh will be 1.6 to 2% less and the fuel component of the price of the generated kWh will be 3 to 4.5% lower. (Z.M.). 5 figs., 3 refs

Cladding failure probability modeling for risk evaluations of fast reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Kramer, J.M.

1987-01-01

This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

International Nuclear Information System (INIS)

1986-09-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

Simulated first operating campaign for the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Benedict, R.W.; Park, K.H.; Ackerman, J.P.

1993-01-01

This report discusses the Integral Fast Reactor (IFR) which is an innovative liquid-metal-cooled reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid-metal cooling to offer significant improvements in reactor safety, operation, fuel cycle-economics, environmental protection, and safeguards. Over the next few years, the IFR fuel cycle will be demonstrated at Argonne-West in Idaho. Spent fuel from the Experimental Breeder Reactor II (EBR-II) win be processed in its associated Fuel Cycle Facility (FCF) using a pyrochemical method that employs molten salts and liquid metals in an electrorefining operation. As part of the preparation for the fuel cycle demonstration, a computer code, PYRO, was developed at Argonne to model the electrorefining operation using thermodynamic and empirical data. This code has been used extensively to evaluate various operating strategies for the fuel cycle demonstration. The modeled results from the first operating campaign are presented. This campaign is capable of processing more than enough material to refuel completely the EBR-II core

Regulations and instructions for RA reactor operation

International Nuclear Information System (INIS)

1989-01-01

This regulatory guide consists of following 4 chapters: Description of the RA reactor, organization scheme, regulations for performing experiments; Regulations for staff on duty; Instructions for operating the vacuum systems, heavy water and helium systems; and evacuation in case of accident [sr

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

Fast reactor operation in the United States

International Nuclear Information System (INIS)

Smith, R.R.; Cissel, D.W.

1978-01-01

Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

N Reactor thermal plume characterization during Pu-only mode of operation

Energy Technology Data Exchange (ETDEWEB)

Ecker, R.M.; Thompson, F.L.; Whelan, G.

1983-04-01

Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

Results of Operation and Utilization of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Le Vinh Vinh; Duong Van Dong; Nguyen Xuan Hai; Pham Ngoc Son; Cao Dong Vu

2014-01-01

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The renovated reactor was put into operation on 20 March 1984. It was designed for the purposes of radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and nuclear education and training. During the last 30 years of operation, the DNRR was efficiently utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine centers and other users in industry, agriculture, hydrology and scientific research; developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples; carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography and nuclear structure study; and establishing nuclear training and education programs for human resource development. This paper presents the results of operation and utilization of the DNRR. In addition, some main reactor renovation projects carried out during the last 10 years are also mentioned in the paper. (author)

Control of water chemistry in operating reactors

International Nuclear Information System (INIS)

Riess, R.

1997-01-01

Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ''modified'' B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs

Control of water chemistry in operating reactors

Energy Technology Data Exchange (ETDEWEB)

Riess, R [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-02-01

Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ``modified`` B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs.

Some particular problems put by operating experimental reactors