Development of self-actuated shutdown system using curie point electromagnet

International Nuclear Information System (INIS)

Kim, Tae Ryong; Park, Jin Ho

1999-01-01

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system (SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet (CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid Metal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design. (author)

Backup passive reactivity shutdown systems

International Nuclear Information System (INIS)

Ashurko, Yu.M.; Kuznetsov, L.A.

1996-01-01

The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs

Backup passive reactivity shutdown systems

Energy Technology Data Exchange (ETDEWEB)

Ashurko, Yu M; Kuznetsov, L A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1996-12-01

The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs.

On line testing of shutdown system

International Nuclear Information System (INIS)

Ramnath, S.; Swaminathan, P.; Sreenivasan, P.

1997-01-01

For ensuring high reliability and availability, safety related Instrumentation channels are triplicated. Solid state electronics can fail in safe or unsafe mode. Hence, it is necessary to supervise the safety related Instrumentation channels from sensor to final shutdown system. Microprocessor/ Microcontroller/ ASIC based online supervision systems are detailed in this paper. (author)

Effect of dc-power-system reliability on reactor-shutdown cooling

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Baranowsky, P.W.; Hickman, J.W.

1981-01-01

The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly with regard to multiple and common cause failures. This concern relates to the application of the single failure criterion for assuring a reliable DC power supply which may be required for the functionability of shutdown cooling systems. The results are presented of a reliability based study performed to assess the adequacy of DC power supply design requirements for currently operating light water reactors with particular attention to shutdown cooling requirements

Nuclear reactor shutdown system

International Nuclear Information System (INIS)

Mangus, J.D.; Cooper, M.H.

1982-01-01

An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

Updating of the program for simulation of Darlington shutdown and regulation systems

International Nuclear Information System (INIS)

1988-07-01

This report describes the current status of the developments of a simulation of the Darlington Nuclear Generating Station shutdown and regulating systems, DARSIM done under contract to the Atomic Energy Control Board (AECB). The DARSIM program simulates the spatial neutron dynamics, the regulation of the reactor power, and shutdown system 1 and shutdown system 2 software. The DARSIM program operates in the interactive simulation program environment. DARSIM was installed on the APOLLO computer at the AECB and a version for an IBM-PC was also provided for the exclusive use of the AECB. Shutdown system software was updated to incorporate the latest revisions in the functional specifications. Additional developments have been provided to assist in the use and interpretation of the DARSIM results

Letter report seismic shutdown system failure mode and effect analysis

International Nuclear Information System (INIS)

KECK, R.D.

1999-01-01

The Supply Ventilation System Seismic Shutdown ensures that the 234-52 building supply fans, the dry air process fans and vertical development calciner are shutdown following a seismic event. This evaluates the failure modes and determines the effects of the failure modes

Fluid shut-down system for a nuclear reactor

International Nuclear Information System (INIS)

Barclay, F.W.; Frey, J.R.; Wilson, J.N.; Besant, R.W.

1975-01-01

A nuclear reactor shut-down system is described which comprises a fluidic vortex valve for releasably maintaining a liquid neutron poison outside of the reactor core, the poison being contained by a reservoir and biased by pressure for flow into poison tubes within the reactor. The upper ends of the poison tubes communicate with the supply port of the vortex valve. A continuous gas flow into the control port maintains normal controlled operation. Shut-down is effected by interruption of the control input. One embodiment comprises three groups of poison tubes and one vortex valve associated with each group wherein shut-down is effected by poison release in two out of the three groups. Preferably, each vortex valve comprises three control ports which operate on a ''voting'' or two-out-of-three basis. (Official Gazette)

Optimal test intervals for shutdown systems for the Cernavoda nuclear power station

International Nuclear Information System (INIS)

Negut, Gh.; Laslau, F.

1993-01-01

Cernavoda nuclear power station required a complete PSA study. As a part of this study, an important goal to enhance the effectiveness of the plant operation is to establish optimal test intervals for the important engineering safety systems. The paper presents, briefly, the current methods to optimize the test intervals. For this reason it was used Vesely methods to establish optimal test intervals and Frantic code to survey the influence of the test intervals on system availability. The applications were done on the Shutdown System no. 1, a shutdown system provided whit solid rods and on Shutdown System no. 2 provided with injecting poison. The shutdown systems receive nine total independent scram signals that dictate the test interval. Fault trees for the both safety systems were developed. For the fault tree solutions an original code developed in our Institute was used. The results, intended to be implemented in the technical specifications for test and operation of Cernavoda NPS are presented

Primary shutdown system monitoring unit for nuclear power plants

International Nuclear Information System (INIS)

Khan, Tahir Kamal; Balasubramanian, R.; Agilandaeswari, K.

2013-01-01

Shut off rods made up of neutron absorbing material are used as Primary Shutdown System. To reduce the power of the reactor under certain abnormal operating conditions, these rods must go down into the core within a specified time. Any malfunctioning in the movement of rods cannot be tolerated and Secondary Shutdown System (SSS) must be actuated within stipulated time to reduce the reactor power. A special safety critical, hardwired electronics unit has been designed to detect failure of PSS Shut off rods movements and generate trip signals for initiating SSS. (author)

Shutdown Chemistry Process Development for PWR Primary System

Energy Technology Data Exchange (ETDEWEB)

Sung, K.B. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

1997-12-31

This study report presents the shutdown chemistry of PWR primary system to reduce and remove the radioactive corrosion products which were deposited on the nuclear fuel rods surface and the outside of core like steam generator channel head, RCS pipings etc. The major research results are the follows ; the deposition radioactive mechanism of corrosion products, the radiochemical composition, the condition of coolant chemistry to promote the dissolution of radioactive cobalt and nickel ferrite, the control method of dissolved hydrogen concentration in the coolant by the mechanical and chemical methods. The another part of study is to investigate the removal characteristics of corrosion product ions and particles by the demineralization system to suggest the method which the system could be operate effectively in shut-down purification period. (author). 19 refs., 25 figs., 48 tabs.

Seismic qualification of SPX1 shutdown systems - tests and calculations

International Nuclear Information System (INIS)

Brochard, D.; Buland, P.

1988-01-01

The SUPERPHENIX 1 shutdown system is composed of two main systems: the Complementary Shutdown System SAC (Systeme d'Arret Complementaire) and the Primary Shutdown System (SCP) (Systeme de Commande Principal). In case of a seismic event, the insertability of the different shutdown systems has to be demonstrated. Tests have been performed on the SAC and have shown that this system was not sensitive to the seismic excitation (the drop time increases of 10% at SSE level). For the SCP, as an analytical demonstration was felt difficult to achieve, it was decided to perform a full scale testing program. These tests have been performed for the two types of SCP which are present in Superphenix: SCP 1 (Creusot Loire design), SCP 2 (Novatome design). As there was no existing facility in France to test this kind of slender structure (21 metres high) a new facility named VESUBIE was designed and installed in an existing pit located at the Saclay nuclear research center. The objectives of the tests were the following: to demonstrate insertability of control rod; to demonstrate absence of seismic induced damage to the SCP; to measure increase of scram time; to measure seismic induced stresses; to obtain data for code correlation. After completion of the tests, measurements have been correlated with results obtained from a non-linear finite element model. Time history correlations were achieved for SCP 1. Afterwards a calculation was performed in hot condition to find if there was some effect of temperature on SCP seismic response. 2 refs, 8 figs

Republic of Korea: Design Study for Passive Shutdown System of the PGSFR

International Nuclear Information System (INIS)

Lee, J.H.

2015-01-01

There have been no experiences of implementing a passive shutdown system in operating or operated SFRs around the world. However, new SFRs are considered to adopt a self-actuated shutdown system (SASS) in the future to provide an alternate means of passively shutting down the reactor. The Prototype Gen-IV SFR (PGSFR) developed by KAERI also adopts this system for the same reason. This passive shutdown design concept is combined with a group of secondary control rod drive mechanisms (SCRDM). The system automatically releases the control rod assembly (CRA) around the set temperature, and then drops the CRA by gravity without any external control signals and any actuating power in an emergency of the reactor. This paper describes the parametric design study of a passive shutdown system, which consists of a thermal expansion device, an electromagnet, and a secondary control rod assembly head. The conceptual design values of each component are also suggested. Parametric calculations are performed to check the suitability of the performance requirements of the thermal expansion device and electromagnets

Evaluation of the safety margins during shutdown for NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Sadek, S.; Bajs, T.

2004-01-01

In the paper the results of RELAP5/mod3.3 calculations of critical parameters during shutdown for NPP Krsko are presented. Conservative evaluations have been performed at NPP Krsko to determine the minimum configuration of systems required for the safe shutdown operation. Critical parameters in these evaluations are defined as the time to start of the boiling and the time of the core dry-out. In order to have better insight into the available margins, the best estimate code RELAP5/mod3.3 has been used to calculate the same parameters. The analyzed transient is the loss of the Residual Heat Removal (RHR) system, which is used to remove decay heat during shutdown conditions. Several configurations that include open and closed Reactor Coolant System (RCS) were considered in the evaluation. The RELAP5/mod3.3 analysis of the loss of the RHR system has been performed for the following cases: 1) RCS closed and water solid, 2) RCS closed and partially drained, 3) Pressurizer manway open, Steam Generator (SG) U tubes partially drained, 4) Pressurizer and SG manways open, SG U tubes completely drained, 5) Pressurizer manway open, SGs drained, SG nozzle dams installed and 6) SG nozzle dams installed, pressurizer manway open, 1 inch break at RHR pump discharge in the loop with pressurizer. Both RHR trains were assumed in operation prior to start of the transient. The maximum average steady state temperature for all analyzed cases was limited to 333 K. (author)

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

International Nuclear Information System (INIS)

Gallardo, J.; Marquino, W.; Mistreanu, A.; Yang, J.

2015-09-01

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

Energy Technology Data Exchange (ETDEWEB)

Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

2015-09-15

The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

Self-actuated shutdown system for a commercial size LMFBR. Final report

Energy Technology Data Exchange (ETDEWEB)

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility and reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power.

Self-actuated shutdown system for a commercial size LMFBR. Final report

International Nuclear Information System (INIS)

Dupen, C.F.G.

1978-08-01

A Self-Actuated Shutdown System (SASS) is defined as a reactor shutdown system in which sensors, release mechanisms and neutron absorbers are contained entirely within the reactor core structure, where they respond inherently to abnormal local process conditions, by shutting down the reactor, independently of the plant protection system (PPS). It is argued that a SASS, having a response time similar to that of the PPS, would so reduce the already very low probability of a failure-to-scram event that costly design features, derived from core disruptive accident analysis, could be eliminated. However, the thrust of the report is the feasibility and reliability of the in-core SASS hardware to achieve sufficiently rapid shutdown. A number of transient overpower and transient undercooling-responsive systems were investigated leading to the selection of a primary candidate and a backup concept. During a transient undercooling event, the recommended device is triggered by the associated rate of change of pressure, whereas the alternate concept responds to the reduction in core pressure drop and requires calibration and adjustment by the operators to accommodate changes in reactor power

Tricon hardware controller implementation of CANDU nuclear power plant shutdown system

International Nuclear Information System (INIS)

Zahedi, P.

2007-01-01

This paper introduces the implementation of logic functions associated with the shutdown systems of CANDU nuclear power plants. The experimental aspects of this work include development of control program embedded in shutdown systems of CANDU based NPPs. A physical test environment is designed to simulate the measurements of in-core flux detector (ICFD) and ion chamber (I/C) signals. The programmable logic used in this experimentation provides Triple Modular Redundant (TMR) architecture as well as a voting mechanism used upon execution of control program on each independent channel. (author)

Pilot Operation of Ex-core Neutron Sensors of Divers Shutdown System (DSS) Unit 2 Ignalina NPP

International Nuclear Information System (INIS)

Jakshtonis, Z.; Krivoshei, G.

2006-01-01

The Ignalina Safety Assessment, which was completed in December 1996, recommended the installation of a diverse shutdown system on the 2nd unit at Ignalina. During the PPR-2004 in the DSS project are created two independent shutdown systems by separating the absorber rods into two independent groups as follows: 1. One system (designated AZ) consists of the existing 24 BAZ rods and 49 AZ/BSM rods that together are used for reliable reactor shutdown (including Control and Protection System (CPS) circuit voiding accident). This system performs the emergency protection function. 2. The other system (designated BSM) comprises the remaining absorber rods and the 49 AZ/BSM rods. Thus 49 AZ/BSM rods are actuated from AZ initiating equipment as well as from BSM initiating equipment. The BSM system performs the normal reactor shutdown function and is able to ensure long-term maintenance of the reactor in the sub-critical state. Along with implementation of DSS was modernized existing Emergency Process Protection System, which was divided into two independent Sets of initiating equipment. The DSS is independent and diverse initiating equipment from the existing 1st Set equipment; with each set having its own independent in-core and ex-core sensors for measurement of neutron flux and process parameters. The 2nd Set of initiating equipment for measuring ex-core neutron flux, was modernized with new design of 4 Ex-Core detectors each have a single low level neutron flux detector and two high range neutron detectors. They are comprising: 1. A fission chamber which operates in pulse mode to cover the low flux levels. 2. A compensated ionisation chamber in current mode to operate at high flux level. This detector is doubled to give a measurement of the axial deviation. Two detectors are enough to produce the axial power deviation. The results of testing and analysis of pilot operation of ex-core neutron sensors of DSS will be shown on the Report. (author)

Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

International Nuclear Information System (INIS)

2015-01-01

A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

The Intelligent Safety System: could it introduce complex computing into CANDU shutdown systems

International Nuclear Information System (INIS)

Hall, J.A.; Hinds, H.W.; Pensom, C.F.; Barker, C.J.; Jobse, A.H.

1984-07-01

The Intelligent Safety System is a computerized shutdown system being developed at the Chalk River Nuclear Laboratories (CRNL) for future CANDU nuclear reactors. It differs from current CANDU shutdown systems in both the algorithm used and the size and complexity of computers required to implement the concept. This paper provides an overview of the project, with emphasis on the computing aspects. Early in the project several needs leading to an introduction of computing complexity were identified, and a computing system that met these needs was conceived. The current work at CRNL centers on building a laboratory demonstration of the Intelligent Safety System, and evaluating the reliability and testability of the concept. Some fundamental problems must still be addressed for the Intelligent Safety System to be acceptable to a CANDU owner and to the regulatory authorities. These are also discussed along with a description of how the Intelligent Safety System might solve these problems

Reactor shutdown back-up system

International Nuclear Information System (INIS)

Hirao, Seizo; Sakashita, Motoaki.

1982-01-01

Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

SNR 2 core dynamics and shut-down signals in a protected loss-of-flow incident

International Nuclear Information System (INIS)

Kleefeldt, K.

1982-01-01

The dynamic behavior of a 1300 MWe Core during a loss-of-flow incident has been analyzed by use of the SAS3D code for a given pump coast down characteristic and constant core inlet temperature. Emphasis was placed on the questions: How fast and via which monitored parameters can the incident be recognized by the reactor protection system. What is the tolerable time span for the shut-down action without exceeding safety limits. Key prameters and limit values as well as conceivable reactivity feed-back effects are discussed. The result is, that three out of four choosen monitored parameters are capable of initiating a shut-down action in time. In addition, the amount of shut-down reactivity required for a successful scram was briefly investigated

Magnetic disconnect for secondary shutdown

International Nuclear Information System (INIS)

Lessor, D.L.

1972-01-01

A description is given of studies to develop a magnetic holding clutch in the control rod drive line as an alternate shutdown device for the FFTF. Results indicate that a three-phase disconnect, hold, and backup shutdown system can be designed to operate satisfactorily. (U.S.)

ORNL Isotopes Facilities Shutdown Program Plan

International Nuclear Information System (INIS)

Gibson, S.M.; Patton, B.D.; Sears, M.B.

1990-10-01

This plan presents the results of a technical and economic assessment for shutdown of the Oak Ridge National Laboratory (ORNL) isotopes production and distribution facilities. On December 11, 1989, the Department of Energy (DOE), Headquarters, in a memorandum addressed to DOE Oak Ridge Operations Office (DOE-ORO), gave instructions to prepare the ORNL isotopes production and distribution facilities, with the exception of immediate facility needs for krypton-85, tritium, and yttrium-90, for safe shutdown. In response to the memorandum, ORNL identified 17 facilities for shutdown. Each of these facilities is located within the ORNL complex with the exception of Building 9204-3, which is located at the Y-12 Weapons Production Plant. These facilities have been used extensively for the production of radioactive materials by the DOE Isotopes Program. They currently house a large inventory of radioactive materials. Over the years, these aging facilities have inherited the problems associated with storing and processing highly radioactive materials (i.e., facilities' materials degradation and contamination). During FY 1990, ORNL is addressing the requirements for placing these facilities into safe shutdown while maintaining the facilities under the existing maintenance and surveillance plan. The day-to-day operations associated with the surveillance and maintenance of a facility include building checks to ensure that building parameters are meeting the required operational safety requirements, performance of contamination control measures, and preventative maintenance on the facility and facility equipment. Shutdown implementation will begin in FY 1993, and shutdown completion will occur by the end of FY 1994

Evolution of shutdown mechanism for PHWRs

International Nuclear Information System (INIS)

Singh, Manjit; Govindarajan, G.

1997-01-01

In 500 MWe PHWR, there are two independent fast acting shutdown systems namely (1) mechanical shut-off rod system and (2) liquid poison injection system. Both systems are independently capable of keeping the reactor in sub-critical condition during long shutdown. Mechanical shut-off rod system being primary shutdown system calls for a very high reliability of operation as well as effectiveness, which are mainly governed by its ability to operate within a very short time and the magnitude of negative reactivity worth it can provide. Mechanical shut-off rods are normally parked above the core by shut-off rod drive mechanism. On receiving a scram signal, shut-off rods are released from the holding electromagnetic clutch and fall under gravity into the core. This paper discusses the salient features of mechanical shut-off rod system. A brief account of detailed design and development of sub-assemblies of shut-off rod drive mechanism is also presented. (author)

Rodded shutdown system for a nuclear reactor

International Nuclear Information System (INIS)

Golden, M.P.; Govi, A.R.

1978-01-01

A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature is described. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core

Inherently safe SNR shutdown system with Curie point controlled sensor/switch unit

International Nuclear Information System (INIS)

Mueller, K.; Norajitra, P.; Reiser, H.

1987-02-01

Inherent shutdown due to increase in the sodium temperature at the core outlet is triggered by interruption of the current supply to the electromagnet coupling of absorber elements via curie point controlled sensor/switch units. These switches are arranged above suitable fuel element positions and spatially independent of the shutdown elements. Compared with other similar systems very short response times are achieved. A prototype switch unit has already undergone extensive testing. These tests have confirmed that switching takes place in a very narrow temperature range. (orig./HP) [de

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

International Nuclear Information System (INIS)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

Energy Technology Data Exchange (ETDEWEB)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance.

Italy: Analysis of Solutions for Passively Actuated Safety Shutdown Devices

International Nuclear Information System (INIS)

Burgazzi, L.

2015-01-01

This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on Fast Reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in Sodium Fast Reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to Fast Reactor design to meet the safety requirements

TRIGA forced shutdowns analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Laslau, Florica

2008-01-01

The need for improving the operation leads us to use new methods and strategies. Probabilistic safety assessments and statistical analysis provide insights useful for our reactor operation. This paper is dedicated to analysis of the forced shutdowns during the first reactor operation period, between 1980 to 1989. A forced shutdown data base was designed using data on forced shutdowns collected from the reactor operation logbooks. In order to sort out the forced shutdowns the records have the following fields: - current number, date, equipment failed, failure type (M for mechanical, E for electrical, D for irradiation device, U for human factor failure; - scram mode, SE for external scram, failure of reactor cooling circuits and/or irradiation devices, SR for reactor scram, exceeding of reactor nuclear parameters, SB for reactor scram by control rod drop, SM for manual scram required by the abnormal reactor status; - scram cause, giving more information on the forced shutdown. This data base was processed using DBase III. The data processing techniques are presented. To sort out the data, one of the criteria was the number of scrams per year, failure type, scram mode, etc. There are presented yearly scrams, total operation time in hours, total unavailable time, median unavailable time period, reactor availability A. There are given the formulae used to calculate the reactor operational parameters. There are shown the scrams per year in the 1980 to 1989 period, the reactor operation time per year, the reactor shutdown time per year and the operating time versus down time per year. Total number of scrams in the covered period was 643 which caused a reactor down time of 4282.25 hours. In a table the scrams as sorted on the failure type is shown. Summarising, this study emphasized some problems and difficulties which occurred during the TRIGA reactor operation at Pitesti. One main difficulty in creating this data base was the unstandardized scram record mode. Some times

Shutdown risk monitoring in TEPCO

International Nuclear Information System (INIS)

Sato, Hiroki; Masuda, Takahiro; Denda, Yasutaka; Yoneyama, Mitsuru; Imai, Shun-ichi; Miyata, Koichi

2009-01-01

At present, we are introducing risk monitors into our all three nuclear power stations; Fukushima Daiichi, Fukushima Daini and Kashiwazaki Kariwa, with technical support of TEPSYS. By monitoring shutdown risk of each unit, we are trying to optimize risks during outage inspection, and raising staff's awareness for reactor safety. This paper presents our recent shutdown risk monitoring activities in Fukushima Daiichi NPS. Shutdown risk monitoring has been carried out for the past five outages of Fukushima Daiichi NPS. Daily-changing shutdown risk is evaluated in the form of core damage frequency (CDF [/day/reactor]). We also examine high-risk point of outage plan if CDF is greater than the threshold at anytime of outage. The results are delivered to operational and maintenance staff before outage. The threshold value is set ten times as much as CDF of unit in operation. As CDF exceeds the threshold, we try to either change the system configuration, or let workers pay more attention to their works during the high-risk period. We already have some examples of outage plan modification to reduce CDF using the risk monitoring information. Greater number of station staff tends to pay more attention to shutdown risk thanks to these activities. (author)

CERN Vacuum-System Activities during the Long Shutdown 1: The LHC Beam Vacuum

CERN Document Server

Baglin, V; Chiggiato, P; Jimenez, JM; Lanza, G

2014-01-01

After the Long Shutdown 1 (LS1) and the consolidation of the magnet bus bars, the CERN Large Hadron Collider (LHC) will operate with nominal beam parameters. Larger beam energy, beam intensities and luminosity are expected. Despite the very good performance of the beam vacuum system during the 2010-12 physics run (Run 1), some particular areas require attention for repair, consolidation and upgrade. Among the main activities, a large campaign aiming at the repair of the RF bridges of some vacuum modules is conducted. Moreover, consolidation of the cryogenic beam vacuum systems with burst disk for safety reasons is implemented. In addition, NEG cartridges, NEG coated inserts and new instruments for the vacuum system upgrade are installed. Besides these activities, repair, consolidation and upgrades of other beam equipment such as collimators, kickers and beam instrumentations are carried out. In this paper, the motivation and the description for such activities, together with the expected beam vacuum performa...

Uncertainty evaluation of reliability of shutdown system of a medium size fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Zeliang, Chireuding; Singh, Om Pal, E-mail: singhop@iitk.ac.in; Munshi, Prabhat

2016-11-15

Highlights: • Uncertainty analysis of reliability of Shutdown System is carried out. • Monte Carlo method of sampling is used. • The effect of various reliability improvement measures of SDS are accounted. - Abstract: In this paper, results are presented on the uncertainty evaluation of the reliability of Shutdown System (SDS) of a Medium Size Fast Breeder Reactor (MSFBR). The reliability analysis results are of Kumar et al. (2005). The failure rate of the components of SDS are taken from International literature and it is assumed that these follow log-normal distribution. Fault tree method is employed to propagate the uncertainty in failure rate from components level to shutdown system level. The beta factor model is used to account different extent of diversity. The Monte Carlo sampling technique is used for the analysis. The results of uncertainty analysis are presented in terms of the probability density function, cumulative distribution function, mean, variance, percentile values, confidence intervals, etc. It is observed that the spread in the probability distribution of SDS failure rate is less than SDS components failure rate and ninety percent values of the failure rate of SDS falls below the target value. As generic values of failure rates are used, sensitivity analysis is performed with respect to failure rate of control and safety rods and beta factor. It is discovered that a large increase in failure rate of SDS rods is not carried to SDS system failure proportionately. The failure rate of SDS is very sensitive to the beta factor of common cause failure between the two systems of SDS. The results of the study provide insight in the propagation of uncertainty in the failure rate of SDS components to failure rate of shutdown system.

Advances in the physics modelling of CANDU liquid injection shutdown systems

International Nuclear Information System (INIS)

Smith, H.J.; Robinson, R.; Guertin, C.

1993-01-01

The physics modelling of liquid poison injection shutdown systems in CANDU reactors accounts for the major phenomena taking place by combining the effects of both moderator hydraulics and neutronics. This paper describes the advances in the physics modelling of liquid poison injection shutdown systems (LISS), discusses some of the effects of the more realistic modelling, and briefly describes the automation methodology. Modifications to the LISS methodology have improved the realism of the physics modelling, showing that the previous methodology significantly overestimated energy deposition during the simulation of a loss of coolant transient in Bruce A, by overestimating the reactivity transient. Furthermore, the automation of the modelling process has reduced the time needed to carry put LISS evaluations to the same level as required for shutoff-rod evaluations, while at the same time minimizing the amount of input, and providing a method for tracing all files used, thus adding a level of quality assurance to the calculation. 5 refs., 11 figs

Shutdown problems in large tokamaks

International Nuclear Information System (INIS)

Weldon, D.M.

1978-01-01

Some of the problems connected with a normal shutdown at the end of the burn phase (soft shutdown) and with a shutdown caused by disruptive instability (hard shutdown) have been considered. For a soft shutdown a cursory literature search was undertaken and methods for controlling the thermal wall loading were listed. Because shutdown computer codes are not widespread, some of the differences between start-up codes and shutdown codes were discussed along with program changes needed to change a start-up code to a shutdown code. For a hard shutdown, the major problems are large induced voltages in the ohmic-heating and equilibrium-field coils and high first wall erosion. A literature search of plasma-wall interactions was carried out. Phenomena that occur at the plasma-wall interface can be quite complicated. For example, material evaporated from the wall can form a virtual limiter or shield protecting the wall from major damage. Thermal gradients that occur during the interaction can produce currents whose associated magnetic field also helps shield the wall

Development and validation of the shutdown cooling system CATHENA model for Gentilly-2

International Nuclear Information System (INIS)

Lecuyer, H.; Hasnaoui, C.; Sabourin, G.; Chapados, S.

2008-01-01

A CATHENA representation of the Gentilly-2 Shutdown Cooling system has been developed for Hydro-Quebec. The model includes the SDCS circuit piping, valves, pumps and heat exchangers. The model is integrated in the G2 CATHENA overall plant model and coupled with the plant control software simulator TROLG2 to allow the simulation of various plant operational modes using the SDCS. Results have been obtained for normal cooling of the primary heat transport system following a planned shut down (transition from full power to shutdown) and for two special SDCS configurations that were used on September 14 and 15, 2006 at Gentilly-2. The results show close match with values measured at Gentilly-2 during either steady or transient states. (author)

Development and validation of the shutdown cooling system CATHENA model for Gentilly-2

Energy Technology Data Exchange (ETDEWEB)

Lecuyer, H.; Hasnaoui, C. [Nucleonex Inc., Westmount, Quebec (Canada); Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Chapados, S. [Hydro-Quebec, Unite Analyse et Fiabilite, Montreal, Quebec (Canada)

2008-07-01

A CATHENA representation of the Gentilly-2 Shutdown Cooling system has been developed for Hydro-Quebec. The model includes the SDCS circuit piping, valves, pumps and heat exchangers. The model is integrated in the G2 CATHENA overall plant model and coupled with the plant control software simulator TROLG2 to allow the simulation of various plant operational modes using the SDCS. Results have been obtained for normal cooling of the primary heat transport system following a planned shut down (transition from full power to shutdown) and for two special SDCS configurations that were used on September 14 and 15, 2006 at Gentilly-2. The results show close match with values measured at Gentilly-2 during either steady or transient states. (author)

Seismic design margin evaluation of systems and equipment required for safe shutdown of North Anna, Units 1 and 2, following an SSE (safe-shutdown earthquake) event. Technical report

International Nuclear Information System (INIS)

Desai, K.D.

1981-06-01

The Advisory Committee on Reactor Safeguards recommended that the NRC staff review in detail the capability and available seismic design margin of fluid systems and equipment used in North Anna, Units 1 and 2 to achieve safe shutdown following a site-design safe-shutdown earthquake (SSE). The staff conducted a series of plant visits and meetings with the licensee to view and discuss the seismic design methodology used for systems, equipment and their supports. The report is a description and evaluation of the seismic design criteria, design conservatisms and seismic design margin for North Anna, Units 1 and 2

Inspection maintenance and planning of shutdown in thermal electric generating plants

International Nuclear Information System (INIS)

Dezordi, W.L.; Correa, D.A.; Kina, M.

1984-01-01

The schedule shutdown of an industrial plant and, more specifically, of an electrical generating station, is becoming increasingly important. The major parameters to be taken into account for the planning of such a shutdown are basically of economic-financial nature such as costs of the related services (materials, equipment, manpower, etc), loss of revenue caused by the station's shutdown as well as by the station availability, and other requirements expected from it by the Load Dispatch and consumers. Improving the equipment's performances and the station's availability are the fundamental objectives to be strived for. The authors present in this paper, in an abridged form, the planning tools used for thermal electric generating plants shutdowns for inspections, maintenance and design changes implementation. (Author) [pt

Reliability modeling of Clinch River breeder reactor electrical shutdown systems

International Nuclear Information System (INIS)

Schatz, R.A.; Duetsch, K.L.

1974-01-01

The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

Nuclear reactor unit shutdown planning

International Nuclear Information System (INIS)

Gardais, J.P.

1994-01-01

In order to optimize the reactor maintenance shutdown efficiency and the reactor availability, an audit had been performed on the shutdown organization at EDF: management, skills, methods and experience feedback have been evaluated; several improvement paths have been identified: project management, introduction of shutdown management professionals, shutdown permanent industrialization, and experience feedback engineering

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

Kinetic analyses on startup and shutdown chemistry of BWR plant

International Nuclear Information System (INIS)

Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

2012-09-01

During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

On line test of trip channels and actuators in primary shutdown system for RAPP-3,4/KAIGA-1,2 reactors

International Nuclear Information System (INIS)

Pramanik, M.; Gupta, P.K.; Ravi Prakash

1997-01-01

Several types of system design and logic arrangements have been used for reactor shutdown systems to avoid the possibility that a single failure within the trip channels/shutdown system actuators can prevent a shutdown system actuation. The trip channels and the logic arrangements associated with the shutdown systems use redundancy to allow them to continue to operate successfully even after having a certain number of failures. A periodic test is thus needed to detect and repair/replace failed elements to prevent accumulation and eventual system failure. The test must be capable of detecting the first failure. The design initiates shutdown system actuation by deenergising the logic relays and turning off the power to the final electrical actuators. Thus, the systems are fail safe with respect to loss of electrical power to the instruments, logic channels and the actuators. Several system/logic arrangements are used to reduce the chances of spurious actuation caused by the loss of a single power supply and other single failures. In general, the systems use coincidence of instrument channel trips and have separate power supplies for the individual instrument channel and dual power supplies where a single final control element is used. These features also permit on line test of instrument channels and logic train. On line test detects component failures not found by other means. The test determines whether gross failure has occurred rather than perform a calibration. As far as practicable the whole channel from sensors to logic and final control element is to be tested. (author)

Certificate for Safe Emergency Shutdown of Wind Turbines

DEFF Research Database (Denmark)

Wisniewski, Rafal; Svenstrup, Mikael; Pedersen, Andreas Søndergaard

2013-01-01

To avoid damage to a wind turbine in the case of a fault or a large wind gust, a detection scheme for emergency shutdown is developed. Specifically, the concept of a safety envelope is introduced. Within the safety envelope, the system can be shutdown without risking structural damage to the turb...

Brief account of the design philosophy for third Qinshan NPP shutdown safety system based on practical application

International Nuclear Information System (INIS)

Xiong Weihua

2005-01-01

Qinshan CANDU power plant is uses the Canadian proven CANDU6 nuclear power technology. It has two characteristic: 1. heavy water-as moderator and coolant; 2. natural uranium as the fuel and change fuel during normal operating. CANDU6 include four special safety system: the No.1 shutdown system (SDS No.1), the No.2 shutdown system (SDS No.2), the containment system, the emergency core cooling system (ECCS). QinShan CANDU power plant is the first commercial PHWR nuclear power plant in China. And some aspect is not similar to everybody. The intention of the article is to introduce the basic design and functions. (authors)

On the speed of response of an FPGA-based shutdown system in CANDU nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

She Jingke, E-mail: jshe2@uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario, N6A 5B9 (Canada); Jiang Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, The University of Western Ontario, London, Ontario, N6A 5B9 (Canada)

2011-06-15

Highlights: > Design and implementation of an FPGA-based CANDU SDS1. > Hardware-in-the-loop simulation for performance evaluation involved with an NPP simulator. > Comparison of the response time between FPGA-based trip channel and software-based PLC. - Abstract: Several issues in an FPGA based implementation of shutdown systems in CANDU nuclear power plants have been investigated in this paper. A particular attention is on the response time of an FPGA implementation of safety shutdown systems in comparison with operating system based software solutions as in existing CANDU plants. The trip decision logic under 'steam generator (SG) level low' condition has been examined in detail. The design and implementation of this logic on an FPGA platform have been carried out. The functionality tests are performed in a hardware-in-the-loop (HIL) environment by connecting the FPGA based system to an NPP simulator, and replacing one channel of Shutdown System Number 1 (SDS1) in the simulator by the FPGA implementation. The response time of the designed system is also measured through multiple tests under different conditions, and statistical data analysis has been performed. The results of the response time tests are compared against those of a software-based implementation of the same trip logic.

Evaluation of slow shutdown system flux detectors in Point Lepreau Generating Station - I: dynamic response characterization

Energy Technology Data Exchange (ETDEWEB)

Anghel, V.N.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Comeau, D. [New Brunswick Power Nuclear, Point Lepreau, New Brunswick (Canada); McKay, J.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Taylor, D. [New Brunswick Power Nuclear, Point Lepreau, New Brunswick (Canada)

2009-07-01

CANDU reactors are protected against reactor overpower by two independent shutdown systems: Shut Down System 1 and 2 (SDS1 and SDS2). At the Point Lepreau Generating Station (PLGS), the shutdown systems can be actuated by measurements of the neutron flux by Platinum-clad Inconel In-Core Flux Detectors (ICFDs). These detectors have a complex dynamic behaviour, characterized by 'prompt' and 'delayed' components with respect to immediate changes in the in-core neutron flux. The dynamic response components need to be determined accurately in order to evaluate the effectiveness of the detectors for actuating the shutdown systems. The amplitudes of the prompt and the delayed components of individual detectors were estimated over a period of several years by comparison of archived detector response data with the computed local neutron flux evolution for SDS1 and SDS2 reactor trips. This was achieved by custom-designed algorithms. The results of this analysis show that the dynamic response of the detectors changes with irradiation, with the SDS2 detectors having 'prompt' signal components that decreased significantly with irradiation. Some general conclusions about detector aging effects are also drawn. (author)

Potential improvement of CANDU NPP safety margins by shortening the response time of shutdown systems using FPGA based implementation

Energy Technology Data Exchange (ETDEWEB)

Jingke She, E-mail: jshe2@uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jin Jiang, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario N6A 5B9 (Canada)

2012-03-15

Highlights: Black-Right-Pointing-Pointer Quantitative analysis of the safety margin improvement through thermalhydraulic simulation and analysis. Black-Right-Pointing-Pointer Hardware-in-the-loop simulation of realizing the improvement by an FPGA-based SDS1. Black-Right-Pointing-Pointer Verification of potential operating power upgrade without endangering the plant safety. - Abstract: The relationship between the peak values of critical reactor variables, such as neutronic power, inside a CANDU reactor and the speed of the response of its shutdown system has been analyzed in the event of a large loss of coolant accident (LOCA). The advantage of shortening the response time of the shutdown action has been demonstrated in term of the improved safety margin. A field programmable gate array (FPGA) platform has been chosen to implement such a shutdown system. Hardware-in-the-loop (HIL) simulations have been performed to demonstrate the feasibility of this concept. Furthermore, connections between the speed of response of the shutdown system and the nominal operating power level of the reactor have been drawn to support for potential power upgrade for existing power plants.

Radiochemical guidelines and process specifications for reactor shutdown: the EDF strategy

International Nuclear Information System (INIS)

Mole, D.; Wintergerst, M.; Meylogan, Th.; Rocher, A.; Sagot, M.J.; Bonelli, V.; Bonnefon, J.; Dupont, B.

2012-09-01

Changes to French nuclear regulations made in June 2006 [1.] have made it necessary for EDF to modify its ruling principles. These modifications required the restructuring of radiochemical guidelines to better reflect their impact on nuclear safety, the environment and radioprotection. In accordance with these aims, a new authoritative document has been produced. This ruling document identifies all parameters with a potential impact on nuclear safety, radiological releases to the environment and personnel dose rates. These diagnostic and control parameters have been identified for a reactor in production and for a reactor during shutdown. For parameters related to a reactor in production, some indicators are used to evaluate impacts on availability, radioprotection and the environment during shutdown and on outage and to anticipate mitigation ways. On the other side, several parameters related to the stages of shutdown were also directly evaluated in order to minimize the impacts. This paper describes the EDF methodology used to establish operational documents: radiochemical guidelines and process specifications, and includes the following: - description of monitored parameters and their associated areas of risk; - justification of target values, frequencies of inspection and the required actions for the monitored parameters. The sizing methodology is based on theoretical studies and on EDF operational experience analysis. By implementing in the operational and technical specifications requirements linked to nuclear safety, radioprotection and environment respect, EDF will benefit from an improved compromise between these areas as well as an increased focus. (authors)

Failure and Reliability Analysis for the Master Pump Shutdown System

International Nuclear Information System (INIS)

BEVINS, R.R.

2000-01-01

The Master Pump Shutdown System (MPSS) will be installed in the 200 Areas of the Hanford Site to monitor and control the transfer of liquid waste between tank farms and between the 200 West and 200 East areas through the Cross-Site Transfer Line. The Safety Function provided by the MPSS is to shutdown any waste transfer process within or between tank farms if a waste leak should occur along the selected transfer route. The MPSS, which provides this Safety Class Function, is composed of Programmable Logic Controllers (PLCs), interconnecting wires, relays, Human to Machine Interfaces (HMI), and software. These components are defined as providing a Safety Class Function and will be designated in this report as MPSS/PLC. Input signals to the MPSS/PLC are provided by leak detection systems from each of the tank farm leak detector locations along the waste transfer route. The combination of the MPSS/PLC, leak detection system, and transfer pump controller system will be referred to as MPSS/SYS. The components addressed in this analysis are associated with the MPSS/SYS. The purpose of this failure and reliability analysis is to address the following design issues of the Project Development Specification (PDS) for the MPSS/SYS (HNF 2000a): (1) Single Component Failure Criterion, (2) System Status Upon Loss of Electrical Power, (3) Physical Separation of Safety Class cables, (4) Physical Isolation of Safety Class Wiring from General Service Wiring, and (5) Meeting the MPSS/PLC Option 1b (RPP 1999) Reliability estimate. The failure and reliability analysis examined the system on a component level basis and identified any hardware or software elements that could fail and/or prevent the system from performing its intended safety function

Shutdown chemistry optimization at Maanshan NPP

International Nuclear Information System (INIS)

Sun Yuanlung; Chuang Benjamin; Su Kouhwa; Kao Jueiting

2009-01-01

At Maanshan PWRs, a significant piping radiation buildup caused by crud burst from fuel surface in the beginning of RFO used to be blamed as a contribution to high personal exposures during outage. Therefore, several modifications on shutdown chemistry procedures such as, early lithium removal, rapid boration, dissolved hydrogen removal, extended RCP operation, and maintaining maximum let down flow, have been consecutively conducted since no.1RFO-16, 2006. The important operational and chemical parameters of modified shutdown chemistry procedures adopted in no.2 RFO-17, 2008 and superiority in low reading (2 mSv/hr) from let down heat exchangers area radiation monitor over 11mSv/hr of no.1 RFO-16 at the same area will be addressed in this paper. At the end of no.2 RFO-17, low personal exposures of 765 man-mSv (TLD)verified the absence of crud burst during shutdown chemistry process and broke records of Maanshan NPP as well. Even with a new job on PZR pre-emptive dissimilar weld overlay which exhausting 17.37% of total 797 man-mSv(TLD) in the latest no.1 RFO-18, 659 man-mSv (TLD) made another record low in the history of Maanshan. (author)

Experience with after-shutdown decay heat removal - BWRs and PWRs

International Nuclear Information System (INIS)

Haugh, J.J.; Mollerus, F.J.; Booth, H.R.

1992-01-01

Boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) make use of residual heat removal systems (RHRSs) during reactor shutdown. RHRS operational events involving an actual loss or significant degradation of an RHRS during shutdown heat removal are often prompted or aggravated by complex, changing plant conditions and by concurrent maintenance operations. Events involving loss of coolant inventory, loss of decay heat removal capability, or inadvertent pressurization while in cold shutdown have occurred. Because fewer automatic protective fetures are operative during cold shutdowns, both prevention and termination of events depend heavily on operator action. The preservation of RHRS cooling should be an important priority in all shutdown operations, particularly where there is substantial decay heat and a reduced water inventory. 13 refs., 3 figs., 4 tabs

Analysis of solutions for passively activated safety shutdown devices for SFR

International Nuclear Information System (INIS)

Burgazzi, Luciano

2013-01-01

Highlights: • Innovative systems for emergency shut down of fast reactors are proposed. • The concepts of inherent and passive safety are put forward. • The relative analysis in terms of safety and reliability is presented. • A comparative assessment among the concepts is performed. • Path forward is tracked. -- Abstract: In order to enhance the inherent safety of fast reactors, innovative reactivity control systems have been proposed for intrinsic ultimate shut-down instead of conventional scram rods, to cope with the potential consequences of severe unprotected transient accidents, such as an energetic core disruptive accident, as in case of sodium fast reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to fast reactor design to meet the safety requirements. This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on fast reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in sodium fast reactors

Preliminary aseismic analysis on bolts of driving mechanism in absorption sphere shutdown system

International Nuclear Information System (INIS)

Chen Feng; Li Tianjin; Zhang Zhengming; Huang Zhiyong; Bo Hanliang

2012-01-01

The absorption sphere shutdown system performs an important role in reactivity regulating and control. Driving mechanism is a set of key mechanical moving parts which is used to control falling of absorption spheres in absorption sphere shutdown system. It is about 5 m for driving mechanism with the slim structure, which is connected with the upper supported plate of metal reactor internals through storage vessel with bolts. Both the storage vessel and driving mechanism are equipment of seismic classification I. It is significant to calculate and check the bolts strength of driving mechanism. In this paper, complicate structure of driving mechanism was simplified to three variable cross sections and statically indeterminate problem was solved. The bolts at the bottom and on the top of the storage vessel were calculated and checked. The preliminary results indicate that the bolts strength is reliable and safe, and the supporting force at the most weak point of driving mechanism is as well obtained. (authors)

Reliability Centered Maintenance (RCM) Methodology and Application to the Shutdown Cooling System for APR-1400 Reactors

Energy Technology Data Exchange (ETDEWEB)

Faragalla, Mohamed M.; Emmanuel, Efenji; Alhammadi, Ibrahim; Awwal, Arigi M.; Lee, Yong Kwan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

Shutdown Cooling System (SCS) is a safety-related system that is used in conjunction with the Main Steam and Main or Auxiliary Feedwater Systems to reduce the temperature of the Reactor Coolant System (RCS) in post shutdown periods from the hot shutdown operating temperature to the refueling temperature. In this paper RCM methodology is applied to (SCS). RCM analysis is performed based on evaluation of Failure Modes Effects and Criticality Analysis (FME and CA) on the component, system and plant. The Logic Tree Analysis (LTA) is used to determine the optimum maintenance tasks. The main objectives of RCM is the safety, preserve the System function, the cost-effective maintenance of the plant components and increase the reliability and availability value. The RCM methodology is useful for improving the equipment reliability by strengthening the management of equipment condition, and leads to a significant decrease in the number of periodical maintenance, extended maintenance cycle, longer useful life of equipment, and decrease in overall maintenance cost. It also focuses on the safety of the system by assigning criticality index to the various components and further selecting maintenance activities based on the risk of failure involved. Therefore, it can be said that RCM introduces a maintenance plan designed for maximum safety in an economical manner and making the system more reliable. For the SCP, increasing the number of condition monitoring tasks will improve the availability of the SCP. It is recommended to reduce the number of periodic maintenance activities.

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

Evaluation of slow shutdown system flux detectors in Point Lepreau Generating Station - II: dynamic compensation error analysis

Energy Technology Data Exchange (ETDEWEB)

Anghel, V.N.P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Taylor, D. [New Brunswick Power Nuclear, Point Lepreau, New Brunswick (Canada)

2009-07-01

CANDU reactors are protected against reactor overpower by two independent shutdown systems: Shut Down System 1 and 2 (SDS1 and SDS2). At the Point Lepreau Generating Station (PLGS), the shutdown systems can be actuated by measurements of the neutron flux from Platinum-clad Inconel In-Core Flux Detectors. These detectors have a complex dynamic behaviour, characterized by 'prompt' and 'delayed' components with respect to immediate changes in the in-core neutron flux. It was shown previously (I: Dynamic Response Characterization by Anghel et al., this conference) that the dynamic responses of the detectors changed with irradiation, with the SDS2 detectors having 'prompt' signal components that decreased significantly. In this paper we assess the implication of these changes for detector dynamic compensation errors by comparing the compensated detector response with the power-to-fuel and the power-to-coolant responses to neutron flux ramps as assumed by previous error analyses. The dynamic compensation error is estimated at any given trip time for all possible accident flux ramps. Some implications for the shutdown system trip set points, obtained from preliminary results, are discussed. (author)

Preliminary Calculations of Shutdown Dose Rate for the CTS Diagnostics System

DEFF Research Database (Denmark)

Klinkby, Esben Bryndt; Nonbøl, Erik; Lauritzen, Bent

2015-01-01

DTU and IST 2 are partners in the design of a collective Thomson Scattering (CTS) diagnostics for ITER through a contract with F4E. The CTS diagnostic utilizes probing radiation of ~60 GHz emitted into the plasma and, using a mirror, collects the scattered radiation by an array of receivers. Having...... on supplying input which affect the system design. Examples include: - Heatloads on plasma facing mirrors and preliminary stress and thermal analysis - Port plug cooling requirements and it's dependence on system design (in particular blanket cut-out) - Shutdown dose-rate calculations (relative analysis...

Safety considerations for research reactors in extended shutdown

International Nuclear Information System (INIS)

2004-01-01

According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

Safety aspects of unplanned shutdowns and trips

International Nuclear Information System (INIS)

1986-05-01

The issue of unplanned shutdowns and trips is receiving increased attention worldwide in view of its importance to plant safety and availability. There exists significant variation in the number of forced shutdowns for nuclear power plants of the same type operating worldwide. The reduction of the frequency of these events will have safety benefits in terms of reducing the frequency of plant transients and the challenges to the safety systems, and the risks of possible incidents. This report provides an insight into the causes of unplanned shutdowns experienced in operating nuclear power plants worldwide, the good practices that have been found effective in minimizing their occurrence, and the measures that have been taken to reduce these events. Specific information on the experiences, approaches and practices of some countries in dealing with this issue is presented in Appendix A

RECAP, Replacement Energy Cost for Short-Term Reactor Plant Shut-Down

International Nuclear Information System (INIS)

VanKuiken, J.C.; Daun, C.J.; Jusko, M.J.

1995-01-01

1 - Description of program or function: RECAP (Replacement Energy Cost Analysis Package) determines the replacement energy costs associated with short-term shutdowns or de-ratings of one or more nuclear reactors. Replacement energy cost refers to the change in generating-system production cost that results from shutting down a reactor. The cost calculations are based on the seasonal, unit-specific cost estimates for 1988-1991 for all 117 nuclear electricity-generating units in the U.S. RECAP is menu-driven, allowing the user to define specific case studies in terms of parameters such as the units to be included, the length and timing of the shutdown or de-rating period, the unit capacity factors, and the reference year for reporting cost results. In addition to simultaneous shutdown cases, more complicated situations, such as overlapping shutdown periods or shutdowns that occur in different years, can be examined through use of a present-worth calculation option. 2 - Method of solution: The user selects a set of units for analysis, defines a shutdown (or de-rating) period, and specifies any planned maintenance outages, delays in unit start-ups, or changes in default capacity factors. The program then determines which seasonal cost numbers to apply, estimates total and daily costs, and makes the appropriate adjustments for multiple outages if they are encountered. The change in production cost is determined from the difference between the total variable costs (variable fuel cost, variable operation and maintenance cost, and purchased energy cost) when the reactor is available for generation and when it is not. Changes in reference-year dollars are based on gross national product (GNP) price deflators or on optional use inputs. Once RECAP has completed the initial cost estimates for a case study (or series of case studies), present-worth analysis can be conducted using different reference-year dollars and discount rates, as specified by the user. The program uses

Multi-unit shutdown due to boiler feedwater chemical excursion

International Nuclear Information System (INIS)

Diebel, M.E.

1991-01-01

Ontario Hydro's Bruce Nuclear Generating Station 'B' consists of four 935 W CANDU units located on the east shore of Lake Huron in the province of Ontario, Canada. On July 25 and 26, 1989 three of the four operating units were shutdown due to boiler feedwater chemical excursions initiated by a process upset in the Water Treatment Plant that provides demineralized make-up water to all four units. The chemicals that escaped from an ion exchange vessel during a routine regeneration very quickly spread through the condensate make-up system and into the boiler feedwater systems. This resulted in boiler sulfate levels exceeding shutdown limits. A total of 260 GWH of electrical generation was unexpectedly made unavailable to the grid at a time of peak seasonal demand. This event exposed several unforeseen deficiencies and vulnerabilities in the automatic demineralized water make-up quality protection scheme, system designs, operating procedures and the ability of operating personnel to recognize and appropriately respond to such an event. The combination of these factors contributed towards turning a minor system upset into a major multi-unit shutdown. This paper provides the details of the actual event initiation in the Water Treatment Plant and describes the sequence of events that led to the eventual shutdown of three units and near shutdown of the fourth. The design inadequacies, procedural deficiencies and operating personnel responses and difficulties are described. The process of recovering from this event, the flushing out of system piping, boilers and the feedwater train is covered as well as our experiences with setting up supplemental demineralized water supplies including trucking in water and the use of rental trailer mounted demineralizing systems. System design, procedural and operational changes that have been made and that are still being worked on in response to this event are described. The latest evidence of the effect of this event on boiler tube

Evolving the JET virtual reality system for delivering the JET EP2 shutdown remote handling tasks

Energy Technology Data Exchange (ETDEWEB)

Williams, Adrian, E-mail: adrian.williams@oxfordtechnologies.co.uk [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, Oxon, OX14 1RJ (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sanders, Stephen [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, Oxon, OX14 1RJ (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Weder, Gerard [Tree-C Technology BV, Buys Ballotstraat 8, 6716 BL Ede (Netherlands); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Bastow, Roger; Allan, Peter; Hazel, Stuart [CCFE, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

2011-10-15

The quality, functionality and performance of the virtual reality (VR) system used at JET for preparation and implementation of remote handling (RH) operations has been progressively enhanced since its first use in the original JET remote handling shutdown in 1998. As preparation began for the JET EP2 (Enhanced Performance 2) shutdown it was recognised that the VR system being used was unable to cope with the increased functionality and the large number of 3D models needed to fully represent the JET in-vessel components and tooling planned for EP2. A bespoke VR software application was developed in collaboration with the OEM, which allowed enhancements to be made to the VR system to meet the requirements of JET remote handling in preparation for EP2. Performance improvements required to meet the challenges of EP2 could not be obtained from the development of the new VR software alone. New methodologies were also required to prepare source, CATIA models for use in the VR using a collection of 3D software packages. In collaboration with the JET drawing office, techniques were developed within CATIA using polygon reduction tools to reduce model size, while retaining surface detail at required user limits. This paper will discuss how these developments have played an essential part in facilitating EP2 remote handling task development and examine their impact during the EP2 shutdown.

Evolving the JET virtual reality system for delivering the JET EP2 shutdown remote handling tasks

International Nuclear Information System (INIS)

Williams, Adrian; Sanders, Stephen; Weder, Gerard; Bastow, Roger; Allan, Peter; Hazel, Stuart

2011-01-01

The quality, functionality and performance of the virtual reality (VR) system used at JET for preparation and implementation of remote handling (RH) operations has been progressively enhanced since its first use in the original JET remote handling shutdown in 1998. As preparation began for the JET EP2 (Enhanced Performance 2) shutdown it was recognised that the VR system being used was unable to cope with the increased functionality and the large number of 3D models needed to fully represent the JET in-vessel components and tooling planned for EP2. A bespoke VR software application was developed in collaboration with the OEM, which allowed enhancements to be made to the VR system to meet the requirements of JET remote handling in preparation for EP2. Performance improvements required to meet the challenges of EP2 could not be obtained from the development of the new VR software alone. New methodologies were also required to prepare source, CATIA models for use in the VR using a collection of 3D software packages. In collaboration with the JET drawing office, techniques were developed within CATIA using polygon reduction tools to reduce model size, while retaining surface detail at required user limits. This paper will discuss how these developments have played an essential part in facilitating EP2 remote handling task development and examine their impact during the EP2 shutdown.

Reactor shutdown device

International Nuclear Information System (INIS)

Inoue, Toyokazu.

1982-01-01

Purpose: To obtain a highly reliable reactor shutdown device capable of checking its function irrespective of the state whether shutdown or operation in a gas-cooled type reactor. Constitution: A hopper is disposed above a guide tube inserted into the reactor core and particulate neutron absorbers are contained in the hopper. An opening for falling particles is disposed to the bottom of the hopper in opposition to the upper end of the guide pipe and the opening is closed by a plug suspended by way of a weld line so as to be capable of dropping. A power source for supplying electrical current to the weld line is disposed. Accordingly, if the current is supplied to the weld line, the line is cut by welding to fall the plug so that the neutron-absorbing particles fall from the opening into the guide pipe to shutdown the reactor, whereby high reliability is obtained for the operation. (Seki, T.)

FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

International Nuclear Information System (INIS)

Pierce, B.F.

1986-07-01

The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

International Nuclear Information System (INIS)

Mattei, J.M.; Bars, G.

1987-11-01

During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

Shutdowns/scrams at BWRs reported under new 1984 LER rule

International Nuclear Information System (INIS)

Mays, G.T.

1985-01-01

Operating experience data from nuclear power plants are essential for safety and reliability analyses. The Licensee Event Reports (LERs), submitted to the NRC by nuclear power plant utilities, contain much of this data. One of the significant aspects of the new LER rule includes the requirement to report all plant shutdowns whereas prior to 1984, not all shutdowns were reported as LERs. This paper reviews the shutdowns and scrams occurring during the first six months of 1984 at BWRs as reported under the new LER rule. The review focused on systems involved, causes, and personnel interactions

Station blackout with failure of wired shutdown system for AHWR

International Nuclear Information System (INIS)

Srivastava, A.; Contractor, A.D.; Chatterjee, B.; Kumar, Rajesh

2015-01-01

Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. This reactor has several advance safety features. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level without primary coolant pumps. Station blackout (SBO) scenario has become very important in aftermath of Fukushima event. The existing reactor has to demonstrate that design features are sufficient to mitigate the scenario whereas the new reactor design are adding specific features to tackle such scenario for prolonged period. The present study demonstrates the design features of AHWR to mitigate the SBO scenario along with failure of wired shutdown system. SBO event leads to feed water pump trip and loss of condenser vacuum which in turn results into loss of feed water and turbine trip on low condenser vacuum signal. Stoppage of steam flow to the turbine and bypass to the condenser lead to bottling up of the system, causing MHT pressure to rise. In the absence of reactor scram, the pressure continues to rise. Isolation Condenser (IC) valve starts opening at a pressure of 7.65 MPa. The pressure continues to rise as IC system is designed for decay heat removal and reactor power is brought down to decay power level through Passive Poison Injection System (PPIS) when the pressure reaches 8.4 MPa. The analysis shows that the event do not lead to undesirable clad surface temperature rise due to reactor trip by PPIS and decay heat removal for prolonged time by IC system. Thermal hydraulic response of different parameters like pressure, temperatures, and flows in MHT system is analyzed for this scenario. Pressure during transient is found to be well below the system pressure criteria of 110% of design pressure. This analysis highlights the design robustness of AHWR. (author)

BWR startup and shutdown activity transport control

Energy Technology Data Exchange (ETDEWEB)

Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

2010-07-01

This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

Controlled shutdown of a fuel cell

Science.gov (United States)

Clingerman, Bruce J.; Keskula, Donald H.

2002-01-01

A method is provided for the shutdown of a fuel cell system to relieve system overpressure while maintaining air compressor operation, and corresponding vent valving and control arrangement. The method and venting arrangement are employed in a fuel cell system, for instance a vehicle propulsion system, comprising, in fluid communication, an air compressor having an outlet for providing air to the system, a combustor operative to provide combustor exhaust to the fuel processor.

Modelling of liquid injection shutdown system (LISS) in ACR-1000

International Nuclear Information System (INIS)

Boubcher, M.; Colton, A.; Donnelly, J.V.

2008-01-01

Modelling of the Liquid Injection Shutdown System (LISS) in the ACR-1000 reactor core must account for the major phenomena that occur following its activation, namely the moderator hydraulics and core neutronics. The former requires modelling of the poison volumes, their time of entry into the reactor, and their propagation into the moderator after emission from the nozzle. The latter requires the reactivity worth of varying volumes and geometries of poisoned moderator fluid in order to simulate the reactivity effect of the injected poison. The time-dependent poison map is generated from hydraulic calculations, and then the neutronics data for standard geometries and concentrations is constructed using DRAGON. (author)

Maintenance, repair and operation (MRO) of shutdown facilities

International Nuclear Information System (INIS)

Kenny, S.

2006-01-01

What level of maintenance does one apply to a shutdown facility? Well it depends on who you ask. Operations staff sees facilities that have completed their useful life cycle as a cost drain while Decommissioning staff sees this as the start of a new life cycle. Based on the decommissioning plan for the particular facility the building could complete another full life cycle while under decommissioning whether it is in storage with surveillance mode or under active decommissioning. This paper will explore how you maintain a facility and systems for many years after its useful life until final decommissioning is completed. When a building is declared redundant, who looks after it until the final decommissioning end state is achieved? At the AECL, Chalk River Labs site the safe shutdown and turnover process is one key element that initiates the decommissioning process. The real trick is orchestrating maintenance, repair and operation plans for a facility that has been poorly invested in during its last years of useful life cycle. To add to that usually shutdowns are prolonged for many years beyond the expected turnover period. During this presentation I will cover what AECL is doing to ensure that the facilities are maintained in a proper state until final decommissioning can be completed. All facilities or systems travel through the same life cycle, design, construction, commissioning, operation, shutdown and demolition. As we all know, nuclear facilities add one more interesting twist to this life cycle called Decommissioning that lands between shutdown and demolition. As a facility nears the shutdown phase, operations staff loose interest in the facility and stop investing in upgrades, repairs and maintenance but continue to invest and focus on maximizing operations. Facility maintenance standards produced by the International Facility Maintenance Association (IFMA) based on a survey done every year state that 2.2% of the total operating costs for the site should be

Novel hybrid Monte Carlo/deterministic technique for shutdown dose rate analyses of fusion energy systems

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Peplow, Douglas E.; Peterson, Joshua L.; Grove, Robert E.

2014-01-01

Highlights: •Develop the novel Multi-Step CADIS (MS-CADIS) hybrid Monte Carlo/deterministic method for multi-step shielding analyses. •Accurately calculate shutdown dose rates using full-scale Monte Carlo models of fusion energy systems. •Demonstrate the dramatic efficiency improvement of the MS-CADIS method for the rigorous two step calculations of the shutdown dose rate in fusion reactors. -- Abstract: The rigorous 2-step (R2S) computational system uses three-dimensional Monte Carlo transport simulations to calculate the shutdown dose rate (SDDR) in fusion reactors. Accurate full-scale R2S calculations are impractical in fusion reactors because they require calculating space- and energy-dependent neutron fluxes everywhere inside the reactor. The use of global Monte Carlo variance reduction techniques was suggested for accelerating the R2S neutron transport calculation. However, the prohibitive computational costs of these approaches, which increase with the problem size and amount of shielding materials, inhibit their ability to accurately predict the SDDR in fusion energy systems using full-scale modeling of an entire fusion plant. This paper describes a novel hybrid Monte Carlo/deterministic methodology that uses the Consistent Adjoint Driven Importance Sampling (CADIS) method but focuses on multi-step shielding calculations. The Multi-Step CADIS (MS-CADIS) methodology speeds up the R2S neutron Monte Carlo calculation using an importance function that represents the neutron importance to the final SDDR. Using a simplified example, preliminary results showed that the use of MS-CADIS enhanced the efficiency of the neutron Monte Carlo simulation of an SDDR calculation by a factor of 550 compared to standard global variance reduction techniques, and that the efficiency enhancement compared to analog Monte Carlo is higher than a factor of 10,000

Failure of PWR-RHRS under cold shutdown conditions: Experimental results from the PKL test facility

International Nuclear Information System (INIS)

Mandl, R.M.; Umminger, K.J.; Logt, J.V.D.

1991-01-01

The Residual Heat Removal System (RHRS) of a PWR is designed to transfer thermal energy from the core after plant shutdown and maintain the plant in cold shutdown or refuelling conditions for extended periods of time. Initial reactor cooling after shutdown is achieved by dissipating heat through the steam generators (SGs) and discharging steam to the condenser by means of the Turbine Bypass System (TBS). When the reactor coolant temperature has dropped to about 160C and pressure has been reduced to 30 bar the RHRS is placed into operation. it reduces the coolant temperature to 50C within 20 hours after shutdown. The time margin for establishing alternate methods of heat removal following a failure of the RHRS depends on the Reactor Coolant System (RCS) temperature, the decay heat rate and the amount of RCS inventory. During some shutdown operations the RCS may be partially drained (e. g. to perform SG inspections). Decreased primary system inventory can significantly reduce the time available to recover the RHRS's function prior to bulk boiling and possible core uncovery. In the PKL test facility, which simulates a 1,300 MWe 4-loop PWR on a scale 1:145, a failure of RHRS under cold shutdown conditions was performed. This presentation gives a brief description of the test facility followed by the test objectives and results of this experiment

Safety shutdown separators

Science.gov (United States)

Carlson, Steven Allen; Anakor, Ifenna Kingsley; Farrell, Greg Robert

2015-06-30

The present invention pertains to electrochemical cells which comprise (a) an anode; (b) a cathode; (c) a solid porous separator, such as a polyolefin, xerogel, or inorganic oxide separator; and (d) a nonaqueous electrolyte, wherein the separator comprises a porous membrane having a microporous coating comprising polymer particles which have not coalesced to form a continuous film. This microporous coating on the separator acts as a safety shutdown layer that rapidly increases the internal resistivity and shuts the cell down upon heating to an elevated temperature, such as 110.degree. C. Also provided are methods for increasing the safety of an electrochemical cell by utilizing such separators with a safety shutdown layer.

Analysis of shutdown and aftercooling cycles of the A-1 nuclear power plant

International Nuclear Information System (INIS)

Mueller, V.; Vopatril, M.

1977-01-01

A new concept is described of the emergency shut-down and after-cooling of the A-1 reactor based on the elimination of pressure shock and minimization of thermal shock. After-cooling is effected by all circulators which had not been defective before shut-down. During shut-down the pumps run at reduced speed. A diesel generator is used as a self-contained power supply. The after-cooling is classified into three types depending on the machinery power consumption, i.e., normal, emergency and super-emergency. The selection of the power supply and the after-cooling conditions proceeds automatically. A mathematical model is described of A-1 reactor behaviour during different accidents requiring the shut-down and after-cooling. Computer programmes are briefly indicated for the analysis of transients in the primary coolant circuit (ZVJE-73-23, SHOCK A-1), for the analysis of transients resulting from a neutron power controller failure or from a circulator failure (HAZARD), for the analysis of after-cooling processes (DENDEL), and programme SAULIS as an auxiliary programme for processing the results and for the print-out of the DENDEL programme. Steady-state parameters before the failure were found as initial conditions for the calculation of transients. The mathematical model was solved using a system of three computer programmes linked by interprogramme communication. The analysis is described of the cooperation of reactor safety circuits and of the automatic equipment for the reduction of thermal shock in the primary coolant circuit, as is the analysis of reactor accidents related to reactor control and to the safety circuits. Theoretical results are compared with experimental values obtained during the experimental A-1 reactor shut-down and after-cooling. The accuracy of the calculated value for the cooling gas temperature at the central and marginal channel outputs is -10 to +15% during the first 30 s of after-cooling. (J.P.)

A Study on the Risk Reduction Effect by MLCS (Mid-loop Level Control System) of EUAPR using the Low-Power and Shutdown PSA Result

International Nuclear Information System (INIS)

Lee, Keunsung; Choi, Sunmi; Kim, Eden

2016-01-01

The EU-APR design has been developed in order to expand and diversify the global nuclear power market of APR1400. For the improvement of shutdown risk for the EUAPR, the mid-loop level control system (MLCS) is considered during mid-loop operation for the EU-APR, which is not incorporated into SKN 3 and 4 (APR1400 Type) in Korea. Commonly, the risk associated with the NPP can be identified through the PSA. Thus, this paper discusses the low power and shutdown (LPSD) risk reduction effect by MLCS using the Low-Power and Shutdown PSA Result. LPSD level 1 PSA models for EU-APR have been developed. The risk reduction effect by MLCS is discussed. Because the loss of shutdown cooling function during mid-loop is one of the most vulnerable events, the MLCS have a significant influence on CDF in LPSD PSA. The shutdown risk of domestic power plants would likely be reduced if the MLCS is adopted in all operating NPPs in Korea during the mid-loop operation. It is expected that this work will contribute to reduce shutdown risk of domestic power plants

Post Fire Safe Shutdown Analysis Using a Fault Tree Logic Model

International Nuclear Information System (INIS)

Yim, Hyun Tae; Park, Jun Hyun

2005-01-01

Every nuclear power plant should have its own fire hazard analysis including the fire safe shutdown analysis. A safe shutdown (SSD) analysis is performed to demonstrate the capability of the plant to safely shut down for a fire in any given area. The basic assumption is that there will be fire damage to all cables and equipment located within a common fire area. When evaluating the SSD capabilities of the plant, based on a review of the systems, equipment and cables within each fire area, it should be determined which shutdown paths are either unaffected or least impacted by a postulated fire within the fire area. Instead of seeking a success path for safe shutdown given all cables and equipment damaged by a fire, there can be an alternative approach to determine the SSD capability: fault tree analysis. This paper introduces the methodology for fire SSD analysis using a fault tree logic model

Test rig overview for validation and reliability testing of shutdown system software

International Nuclear Information System (INIS)

Zhao, M.; McDonald, A.; Dick, P.

2007-01-01

The test rig for Validation and Reliability Testing of shutdown system software has been upgraded from the AECL Windows-based test rig previously used for CANDU6 stations. It includes a Virtual Trip Computer, which is a software simulation of the functional specification of the trip computer, and a real-time trip computer simulator in a separate chassis, which is used during the preparation of trip computer test cases before the actual trip computers are available. This allows preparation work for Validation and Reliability Testing to be performed in advance of delivery of actual trip computers to maintain a project schedule. (author)

Plasma shutdown device

International Nuclear Information System (INIS)

Hosogane, Nobuyuki; Nakayama, Takahide.

1985-01-01

Purpose: To prevent concentration of plasma currents to the plasma center upon plasma shutdown in a torus type thermonuclear device by the injection of fuels to the plasma center thereby prevent plasma disruption at the plasma center. Constitution: The plasma shutdown device comprises a plasma current measuring device that measures the current distribution of plasmas confined within a vacuum vessel and outputs a control signal for cooling the plasma center when the plasma currents concentrate to the plasma center and a fuel supply device that supplies fuels to the plasma center for cooling the center. The fuels are injected in the form of pellets into the plasmas. The direction and the velocity of the injection are set such that the pellets are ionized at the center of the plasmas. (Horiuchi, T.)

An analysis of multiple particle settling for LMR backup shutdown systems

International Nuclear Information System (INIS)

Brock, R.W.

1992-05-01

Backup shutdown systems proposed for future LMRs may employ discreet absorber particles to provide the negative reactivity insertion. When actuated, these systems release a dense packing of particles from an out-of-core region to settle into an in-core region. The multiple particle settling behavior is analyzed by the method of continuity waves. This method provides predictions of the dynamic response of the system including the average particle velocity and volume fraction of particles vs. time. Although hindered settling problems have been previously analyzed using continuity wave theory, this application represents an extension of the theory to conditions of unrestrained settling. Typical cases are analyzed and numerical results are calculated based on a semi-empirical drift-flux model. For 1/4-inch diameter boron-carbide particles in hot liquid sodium, the unrestrained settling problem assumes a steady-state solution when the average volume fraction of particles is 0.295 and the average particle velocity is 26.0 cm/s

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Wade, D.C.

1988-01-01

The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

Design criteria for a self-actuated shutdown system to ensure limitation of core damage

International Nuclear Information System (INIS)

Deane, N.A.; Atcheson, D.B.

1981-09-01

Safety-based functional requirements and design criteria for a self-actuated shutdown system (SASS) are derived in accordance with LOA-2 success criteria and reliability goals. The design basis transients have been defined and evaluated for the CDS Phase II design, which is a 2550 MWt mixed oxide heterogeneous core reactor. A partial set of reactor responses for selected transients is provided as a function of SASS characteristics such as reactivity worth, trip points, and insertion times

40 CFR 52.271 - Malfunction, startup, and shutdown regulations.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 3 2010-07-01 2010-07-01 false Malfunction, startup, and shutdown..., startup, and shutdown regulations. (a) The following regulations are disapproved because they would permit... malfunctions and/or fail to sufficiently limit startup and shutdown exemptions to those periods where it is...

Requirements Analysis Study for Master Pump Shutdown System Project Development Specification

International Nuclear Information System (INIS)

BEVINS, R.R.

2000-01-01

This study is a requirements document that presents analysis for the functional description for the master pump shutdown system. This document identifies the sources of the requirements and/or how these were derived. Each requirement is validated either by quoting the source or an analysis process involving the required functionality, performance characteristics, operations input or engineering judgment. The requirements in this study apply to the first phase of the W314 Project. This document has been updated during the definitive design portion of the first phase of the W314 Project to capture additional software requirements and is planned to be updated during the second phase of the W314 Project to cover the second phase of the project's scope

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C System

International Nuclear Information System (INIS)

Jang, Seung Cheol; Kang, Hyun Gook; Lim, Ho Gon; Park, Jin Hee; Kang, Dae Il; Eom, Heung Sub; Kim, Man Cheol; Lee, Ho Joong; Kim, Jae Whan; Ha, Jae Joo

2007-06-01

There are two technical areas to deal with in the project: the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: The LPSD PSA Area Ο Quality improvement of the KSNP LPSD PSA model in the following four technical areas; human reliability analysis (HR), system analysis (SY), data analysis (DA) and accident sequence quantification (QU) Ο Development of the LPSD configuration risk management(CRM) model - Study on the methodology for developing a CRM model, so-called ASLOC (Autonomous Shutdown LOgic Creation) - Development of the LPSD CRM model for the units of Ulchin 3 and 4 The Digital I and C PSA Area Ο Development of impact model of ESF-CCS on plant risks - Unavailability analysis of ESF-CCS for APR-1400 - Digital plant risk models for evaluating core damage frequency (CDF) Ο Study on the methodologies for treating digital-specific problems in the digital I and C PSA - Study on the methodology for evaluating safety-critical SW reliability by BBN techniques, including a feasibility study of reliability growth model - Study on the methodology for the safety-critical network system by Markov chain

Operating experiences of reactor shutdown system at MAPS

International Nuclear Information System (INIS)

Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

1997-01-01

The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

Oak Ridge Research reactor shutdown maintenance and surveillance

International Nuclear Information System (INIS)

Coleman, G.H.; Laughlin, D.L.

1991-05-01

The Department of Energy ordered the Oak Ridge Research Reactor to be placed in permanent shutdown on July 14, 1987. The paper outlines routine maintenance activities and surveillance tests performed April through September, 1990, on the reactor instrumentation and controls, process system, and the gaseous waste filter system. Preparations are being made to transfer the facility to the Remedial Action Program. 6 tabs

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Fermi Research Alliance (FRA), Batavia, IL (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-09-30

important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 13 shutdown sites use designs from 4 different suppliers involving 11 different (horizontal and vertical) dry storage systems that would require the use of 9 different transportation cask designs to remove the SNF and GTCC waste from the shutdown sites. Although some changes to transportation certificates of compliance will be required, the SNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in dual purpose dry storage canisters that can be transported, including a small amount of high-burnup fuel. Most sites indicated that 2-3 years of advance time would be required for its preparations before shipments could begin. Some sites could be ready in less time. As additional sites such as Fort Calhoun, Clinton, Quad Cities, Pilgrim, Oyster Creek, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

Design and analysis of shutdown mechanisms of PFBR

International Nuclear Information System (INIS)

Vijayashree, R.; Rajan Babu, V.; Puthiyavinayagam, P.; Chellapandi, P.; Chetal, S.C.

2009-01-01

Prototype Fast Breeder Reactor (PFBR) is equipped with two independent, fast acting and diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR) and that of the second system is called Diverse Safety Rod (DSR). The respective drive mechanisms are called Control and Safety Rod Drive Mechanism (CSRDM) and Diverse Safety Rod Drive Mechanism (DSRDM). The conceptual features of the Absorber Rods (ARs) and Absorber Rod Drive Mechanisms (ARDMs) are given in the figures. The functions and design specifications of the ARDMs are listed. The theoretical results of the performance of the shutdown systems during scram are presented. The design was always backed up with testing and design validation. The individual subassemblies testing and the design have proceeded side by side, the efforts finally culminated into the manufacturing of 1:1 scale prototype ARDMs and ARs. The prototypes were extensively tested in air, water and sodium to qualify them for reactor application. A companion paper in this conference gives the details of design validation by testing. This paper gives a brief account of the design of ARDMs and ARs. (author)

Risks Associated with Shutdown in PWRs

International Nuclear Information System (INIS)

Grlicarev, I.

1996-01-01

The selected set of risks associated with reactor shutdown in PWRs are outlined and discussed (e. g. outage planning, residual heat removal capability, rapid boron dilution, containment integrity, fire protection). The contribution of different outage strategies to overall core damage risk during shutdown is assessed for a particular basic outage plan. The factors which increase or minimize the probability of reactor coolant boiling or core damage are analysed. (author)

Lithium Hideout and Return in the CANDU Heat Transport System during Shutdown and Start-up

International Nuclear Information System (INIS)

Qiu, L.; Snaglewski, A.P.

2012-09-01

Lithium hydroxide is used to control the pH a (pH apparent) of the Heat Transport System (HTS) coolant in CANDU R reactors. The recommended range of the lithium concentration in the coolant is between 0.38 ppm (5.5x10 -5 m) and 0.60 ppm (8.7x10 -5 m) to minimize carbon steel corrosion in the HTS and magnetite deposition in the core during normal operation; this corresponds to pH a values between 10.2 and 10.4. Similar pH a and lithium concentrations should be maintained during shutdown and start-up. However, maintaining the pH a of the HTS coolant within specification during shutdown and start-up has been difficult for some CANDU stations, especially when the HTS is taken to a Low Level Drain State (LLDS), because of lithium hideout and return. This paper presents the results from lithium adsorption and desorption studies on iron oxides under relevant shutdown and start-up chemistry conditions performed to elucidate the mechanisms of the observed lithium hideout and return. The results show that lithium hideout and return are driven largely by changes in the solubility of magnetite as the HTS coolant chemistry changes during shutdown; changes in lithium concentration were inversely correlated with the solubility of magnetite. When the HTS system is de-pressurized and drained to a low coolant level, the ingress of air rapidly oxidizes the dissolved Fe (II) in the coolant, 2Fe +2 + 1 / 2 O 2 + 3 H 2 = 2FEOOH + 4 H + , resulting in the formation of lepidocrocite or maghemite, which have much lower solubilities but larger surface areas than does magnetite. The large surface area of the Fe (III) oxides can adsorb significant quantities of lithium from the coolant, leading to lithium hideout and a pH a decrease. During start-up, the chemistry of the coolant changes from oxidizing to reducing, and lepidocrocite and other Fe (III) oxides are reduced to Fe (II), gradually dissolving as their solubility increases with increasing temperature. The adsorbed lithium is released

Evolution of the ATLAS Distributed Computing system during the LHC Long shutdown

CERN Document Server

Campana, S; The ATLAS collaboration

2014-01-01

The ATLAS Distributed Computing project (ADC) was established in 2007 to develop and operate a framework, following the ATLAS computing model, to enable data storage, processing and bookkeeping on top of the WLCG distributed infrastructure. ADC development has always been driven by operations and this contributed to its success. The system has fulfilled the demanding requirements of ATLAS, daily consolidating worldwide up to 1PB of data and running more than 1.5 million payloads distributed globally, supporting almost one thousand concurrent distributed analysis users. Comprehensive automation and monitoring minimized the operational manpower required. The flexibility of the system to adjust to operational needs has been important to the success of the ATLAS physics program. The LHC shutdown in 2013-2015 affords an opportunity to improve the system in light of operational experience and scale it to cope with the demanding requirements of 2015 and beyond, most notably a much higher trigger rate and event pileu...

Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul

2013-04-30

The Blue Ribbon Commission on America’s Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

FPGA Implementation of the stepwise shutdown system

International Nuclear Information System (INIS)

Lotjonen, L.

2012-01-01

This report elaborates the design process of applications for field-programmable gate array (FPGA) devices. Brief introductions to EPGA technology and the design process are first given and then the design phases are walked through with the aid of a case study. FPGA is a programmable logic device that is programmed by the customer rather than the manufacturer. They are also usually re-programmable which enables updating their programming and otherwise modifying the design. There are also one-time programmable FPGAs that can be used when security issues require it. FPGA is said to be 'hardware designed like software', which means that the design process resembles software development but the end-product is considered a hardware application because the execution of the functions is entirely different from a microprocessor. This duality can give both the flexibility of software and the reliability of hardware. The FPGA design and verification and validation (V and V) methods for NPP safety systems have not yet matured because the technology is rather new in the field. Software development methods and standards can be used to some extent but the hardware aspects bring new challenges that cannot be tackled using purely software methods. International efforts are being made to development formal and consistent design and V and V methodology regulations for FPGA devices. A preventive safety function called Stepwise Shutdown System (SWS) was implemented on an Actel M1 IGLOO field-programmable gate array (FPGA) device. SWS is used to drive a process into a normal state if the process measurements deviate from the desired operating values. This can happen in case of process disturbances. The SWS implementation process from the requirements to the functional device is elaborated. The design is tested via simulation and hardware testing. The case study is to be further expanded as a part of a master's thesis. (orig.)

FPGA Implementation of the stepwise shutdown system

Energy Technology Data Exchange (ETDEWEB)

Lotjonen, L.

2012-07-01

This report elaborates the design process of applications for field-programmable gate array (FPGA) devices. Brief introductions to EPGA technology and the design process are first given and then the design phases are walked through with the aid of a case study. FPGA is a programmable logic device that is programmed by the customer rather than the manufacturer. They are also usually re-programmable which enables updating their programming and otherwise modifying the design. There are also one-time programmable FPGAs that can be used when security issues require it. FPGA is said to be 'hardware designed like software', which means that the design process resembles software development but the end-product is considered a hardware application because the execution of the functions is entirely different from a microprocessor. This duality can give both the flexibility of software and the reliability of hardware. The FPGA design and verification and validation (V and V) methods for NPP safety systems have not yet matured because the technology is rather new in the field. Software development methods and stanfards can be used to some extent but the hardware aspects bring new challenges that cannot be tacled using purely software methods. International efforts are being made to development formal and consistent design and V and V methodology regulations for FPGA devices. A preventive safety function called Stepwise Shutdown System (SWS) was implemented on an Actel M1 IGLOO field-programmable gate array (FPGA) device. SWS is used to drive a process into a normal state if the process measurements deviate from the desired operating values. This can happen in case of process disturbances. The SWS implementation processfrom the reguirements to the functional device is elaborated. The design is tested via simulation and hardware testing. The case study is to be further expanded as a part of a master's thesis. (orig.)

First LHC Shutdown: Coordination and Schedule Issues

CERN Document Server

Coupard, J; Grillot, S

2010-01-01

The first LHC shutdown started in fall 2008, just after the incident on the 19th of September 2008. In addition to the typical work of a shutdown, a large number of interventions, related to the “consolidation after the incident” were performed in the LHC loop. Moreover the amount of work increased during the shutdown, following the recommendations and conclusions of the different working groups in charge of the safety of the personnel and of the machine. This paper will give an overview of the work performed, the organization of the coordination, emphasizing the new safety risks (electrical and cryogenic), and how the interventions were implemented in order to ensure both the safety of personnel and a minimized time window.

LHC Detector Vacuum System Consolidation for Long Shutdown 1 (LS1) in 2013-2014

CERN Document Server

Gallilee, M; Cruikshank, P; Gallagher, J; Garion, C; Jimenez, J M; Kersevan, R; Kos, H; Leduc, L; Lepeule, P; Provot, N; Rambeau, H; Veness, R

2012-01-01

The LHC has ventured into unchartered territory for Particle Physics accelerators. A dedicated consolidation program is required between 2013 and 2014 to ensure optimal physics performance. The experiments, ALICE, ATLAS, CMS, and LHCb, will utilise this shutdown, along with the gained experience of three years of physics running, to make optimisations to their detectors. New vacuum technologies have been developed for the experimental areas, to be integrated during this first phase shutdown. These technologies include bellows, vacuum chambers and ion pumps in aluminium, new beryllium vacuum chambers, and composite mechanical supports. An overview of this first phase consolidation program for the LHC experiments is presented.

40 CFR 63.4768 - What are the requirements for continuous parameter monitoring system installation, operation, and...

Science.gov (United States)

2010-07-01

... device to the atmosphere. (ii) Car-seal or lock-and-key valve closures. Secure any bypass line valve in the closed position with a car-seal or a lock-and-key type configuration. You must visually inspect...) Automatic shutdown system. Use an automatic shutdown system in which the coating operation is stopped when...

Technical Assessment: WRAP 1 HVAC Passive Shutdown

International Nuclear Information System (INIS)

Ball, D.E.; Nash, C.R.; Stroup, J.L.

1993-01-01

As the result of careful interpretation of DOE Order 6430.lA and other DOE Orders, the HVAC system for WRAP 1 has been greatly simplified. The HVAC system is now designed to safely shut down to Passive State if power fails for any reason. The fans cease functioning, allowing the Zone 1 and Zone 2 HVAC Confinement Systems to breathe with respect to atmospheric pressure changes. Simplifying the HVAC system avoided overdesign. Construction costs were reduced by eliminating unnecessary equipment. This report summarizes work that was done to define the criteria, physical concepts, and operational experiences that lead to the passive shutdown design for WRAP 1 confinement HVAC systems

Safe shutdown analysis for submerged equipment inside containment

International Nuclear Information System (INIS)

Song, Dong Soo; Lee, Seung Chan; Yoon, Duk Joo; Ha, Sang Jun

2017-01-01

The purpose of the paper is to analyze internal flooding effects on the submerged safety-related components inside containment building. Safe shutdown analysis has been performed based on the criteria, assumptions and guideline provided in ANSI/ANS-56.11-1988 and ANSI/ANS-58.11-1988. Flooding can be postulated from a failure of several systems located inside the containment. Loss of coolant accident (LOCA), Feed water line break (FWLB), and other pipe breaks/cracks are assumed. The worst case flooding scenario is a large break LOCA. The maximum flood level for a large break LOCA is calculated based on the combined inventory of the reactor coolant system, the three accumulators, the boron injection tank (BIT), the chemical additive tank (CAT), and the refueling water storage tank (RWST) flooding the containment. The maximum flood level that could occur from all of the water which is available in containment is 2.3 m from the base elevation. A detailed flooding analysis for the components has been performed to demonstrate that internal flooding resulting from a postulated initiating event does not cause the loss of equipment required to achieve and maintain safe shutdown of the plant, emergency core cooling capability, or equipment whose failure could result in unacceptable offsite radiological consequences. The flood height can be calculated as h = (dh/dt) x (t-t 0 ) + h 0 , where h = time dependent flood height and subscript 0 means the initial value and height slope dh/dt. In summary, the submerged components inside containment are acceptable because they complete the mission of safety injection (SI) prior to submeregency or have no safe shutdown function including containment isolation during an accident. (author)

Criteria for remote shutdown for light water reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This Standard provides design criteria which require that: (1) specific controls and monitoring equipment shall be provided for achieving and maintaining the plant in a safe shutdown condition; (2) these controls be installed at a location (or locations) that is physically remote from the control room and cable spreading areas; (3) simultaneous control from both locations shall be prevented by administrative controls or devices for transfer of control from the control room to the remote location(s); and (4) the remote controls be used as defense-in-depth measure in addition to the control room shutdown controls and as a minimum shall provide for one complete channel of shutdown equipment

Report on the use of programmable digital computers in the shutdown systems of the Darlington G.S

International Nuclear Information System (INIS)

1981-06-01

This report considers the use of large, programmable computers in the shutdown system at the planned Darlington Generating Station. After a review of the document submitted to the Atomic Energy Control Board (AECB) by Atomic Energy of Canada Ltd. Engineering Company in support of gaining AECB's agreement in principle to such a system, the ACNS concludes that no fundamental principle will be challenged by the introduction of computer technology into safety systems. It cautions, however, that the AECB should ensure that existing principles particularly those of independence and redundancy, will not be compromised in the application

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Federal Railroad Administration (FRA) (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2014-10-01

visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

LHC Report: The shutdown work nearing completion

CERN Multimedia

CERN Bulletin

2011-01-01

The work planned for the LHC injector chain during the winter shutdown is nearing completion. The PS Booster (PSB) and PS will be closed to access next week, and the control of machine access will be transferred to the CERN Control Centre in preparation for the resumption of machine operation. Hardware tests are being performed in all the machines.   Tests are under way in the LHC tunnel. The technical teams are putting the finishing touches to the work planned for the winter shutdown. At the Linac2, the PS Booster and the PS, work will be completed next week and hardware tests will be carried out soon after. POPS, the new powering system for the PS, will be commissioned for the first time in the coming days after the necessary preliminary tests have been carried out. At the SPS, various magnets have been replaced over recent weeks and the performance tests on the main power supply and other hardware tests will be able to start shortly. After that, the machine will be ready for operation with b...

Perspectives on Low Power and Shutdown Risk

International Nuclear Information System (INIS)

Camp, Allen L.; Whitehead, Donnie W.; Wheeler, Timothy A.; Lehner, John; Chu, Tsong-Lun; Lois, Erasmai; Drouin, Mary

2000-01-01

This paper presents results from a program sponsored by the US Nuclear Regulatory Commission to examine the risks from low power and shutdown operations. Significant progress has been made by the industry in reducing such risks; however, important operational events continue to occur. Current perceptions of low power and shutdown risks are discussed in the paper along with an assessment of the current methods for understanding important events and quantifying their associated risk

Comparison of Qualitative and Quantitative Risk Results for Shutdown Operation

International Nuclear Information System (INIS)

Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro; Hong, Sung Yull

2006-01-01

The Defense-In-Depth philosophy is a fundamental concept of nuclear safety. The objective of Defense-In- Depth (DID) evaluation is to assess the level of Defense- In-Depth maintained during the various plant maintenance activities. Especially for shutdown and outage operations, the Defense-In-Depth might be challenged due to the reduction in redundancy and diversity resulting from the maintenance. The qualitative defense-in-depth evaluation using deterministic trees such as SFAT (Safety Function Assessment Tree), can provide 'Safety' related information on the levels of defense-in-depth according to the plant configuration including the levels of redundancy and diversity. For the more reasonable color decision of SFAT, it is necessary to identify the risk impact of degradation of redundancy and diversity of mitigation systems. The probabilistic safety analysis for the shutdown status can provide risk information related on the degradation of redundancy and diversity level for the safety functions during outage. Insights from the both methods for the plant status can be the same or different. The results of DID approach and PSA for the shutdown state are compared in this paper

Thermosyphon Phenomenon as an alternate heat sink of Shutdown Cooling System for the CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Jonghyun [GNEST, Seoul (Korea, Republic of); Lee, Kwangho; Oh, Haechol; Jun, Hwangyong [KEPRI, Taejon (Korea, Republic of)

2006-07-01

During the outage(overhaul) of the CANDU plant, there is a period when the coolant is partially drained to the reactor header level and the coolant is cooled and depressurized by Shutdown Cooling System(SDCS) other than PHTS pump. In the postulated accident of the loss of SDCS-the PHTS pump failure, the primary coolant system should be cooled by the alternate heat sink using the thermosyphon pheonomenon(TS) through the steam generator(SG) This study was aimed at verification and analyzing the core cooling ability of the TS. And the sensitivity analysis was done for the number of SGs used in the TS. As an analysis tool, RELAP5/CANDU was used.

Document status for 1 and 2 Kozloduy NPP decommissioning activities -Phase 'Final Shutdown'

International Nuclear Information System (INIS)

Vangev, A.; Boyadjiev, Z.

1997-01-01

Decommissioning process (D and D) is the final phase of each nuclear reactor life cycle. The first nuclear reactor generation has reached his expiration life date. Decommissioning working documentation had not been taken into account at the project and construction stage. The decommissioning activities, planning and legislation has to develop along their operation. Most of developed nuclear energetic countries have gathered good experience and have create their own decommissioning strategy. This report represents in brief an overview of different country's approaches and the Kozloduy NPP decommissioning activity intention in near future and reviews the D and D working document status for 1 and 2 Kozloduy NPP Units decommissioning. Kozloduy NPP D and D task to the moment is to plan the first stage of the decommissioning process - 'The Final Shutdown' and to prepare the working documents for the phase execution. The Final Shutdown of Kozloduy NPP - 1 is the termination of operation of the Units 1 and 2 and the electricity production cessation after their useful life exhaust. In accordance with the legal legislation in Bulgaria only the normal planned termination of operation on units 1 and 2 should be prescribed. The project results concern the initial condition of the equipment and systems, their preparation and sequence for defueling, decontamination and dismantling. A plan for activities' organization for D and D and Complex Characterization of the Site under consideration will contain the following documents: 1. Time-schedule for the sequence of activities during the stages of the Final Shutdown and Safe Enclosure preparation. Technical project for organization of work related to Final Shutdown; 2. Complex Characterization Programme for a condition investigation of the Units 1 and 2 equipment and systems. 3. Technical project for design modifications and dismantling of equipment and systems which violate the radiation and nuclear safety during the Final Shutdown

Prevention device for rapid reactor core shutdown in BWR type reactors

International Nuclear Information System (INIS)

Koshi, Yuji; Karatsu, Hiroyuki.

1986-01-01

Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

Optimization of reactor coolant shutdown chemistry practices for crud inventory management

International Nuclear Information System (INIS)

Fellers, B.; Barnette, J.; Stevens, J.; Perkins, D.

2002-01-01

This report describes reactor coolant shutdown chemistry control practices at Comanche Peak Steam Electric Station (CPSES, TXU-Generation, USA). The shutdown evolution is managed from a process control perspective to achieve conditions most favorable to crud decomposition and to avoiding re-precipitation of metals. The report discusses the evolution of current industry practices and the necessity for greater emphasis on shutdown chemistry control in response to Axial Offset Anomaly and growth of ex-core radiation fields during outage conditions. Nuclear Industry experience with axial offset anomaly (AOA), radiation field growth and unexpected behavior of crud during reactor shutdowns has encouraged the refinement of chemistry control practices during plant shutdown and startup. The strong implication of nickel rich crud as a cause of AOA and unexpected crud behavior has resulted in a focus on nickel inventory management. The goals for Comanche Peak Steam Electric Station (CPSES) include maintaining solubility of metals and radioisotopes, maximizing nickel removal and effective cleanup with demineralizers. This paper provides results and lessons learned from long term efforts to optimize the shutdown process. (authors)

Startup, Shutdown, & Malfunction (SSM) Emissions

Science.gov (United States)

EPA issued a final action to ensure states have plans in place that are fully consistent with the Clean Air Act and recent court decisions concerning startup, shutdown and malfunction (SSM) operations.

An analysis on water hammer in liquid injection shutdown system of CANDU-9

International Nuclear Information System (INIS)

Kim, T. H.; Heo, J.; Han, S. K.; Choi, H. Y.; No, T. S.

2000-01-01

The water hammer analysis code, PTRAN, is used for computation of transient pressures and pressure differentials in the Liquid Injection Shutdown System(LISS) piping network of CANDU-9 to ensure that the design allowables for LEVEL C Service Limit are met for the water hammer loads resulting from the water hammer. The LISS piping network of CANDU-9 has incorporated design improvement in considering the water hammer, such as declining the horizontal part of helium header, and raising the elevation of the overall system piping configuration, etc. The maximum pressure in the LISS piping network is found to be 7.92 MPa(a) at the closed valve in the vent line, which is below the allowable working pressure and the valve design pressure under Level C service conditions. And it is also shown that the maximum pressure in CANDU-9 is much lower than that in CANDU-6

Stabilization and shutdown of Oak Ridge National Laboratory's Radioisotopes Production Facility

International Nuclear Information System (INIS)

Eversole, R.E.

1992-01-01

The Oak Ridge National Laboratory (ORNL) has been involved in the production and distribution of a variety of radioisotopes for medical, scientific and industrial applications since the late 1940s. Production of these materials was concentrated in a number of facilities primarily built in the 1950s and 1960s. Due to the age and deteriorating condition of these facilities, it was determined in 1989 that it would not be cost effective to upgrade these facilities to bring them into compliance with contemporary environmental, safety and health standards. The US Department of Energy (DOE) instructed ORNL to halt the production of isotopes in these facilities and maintain the facilities in safe standby condition while preparing a stabilization and shutdown plan. The goal was to place the former isotope production facilities in a radiologically and industrially safe condition to allow a 5-year deferral of the initiation of environmental restoration (ER) activities. In response to DOE's instructions, ORNL identified 17 facilities for shutdown, addressed the shutdown requirements for each facility, and prepared and implemented a three-phase, 4-year plan for shutdown of the facilities. The Isotopes Facilities Shutdown Program (IFSP) office was created to execute the stabilization and shutdown plan. The program is entering its third year in which the actual shutdown of the facilities is initiated. Accomplishments to date have included consolidation of all isotopes inventory into one facility, DOE approval of the IFSP Environmental Assessment (EA), and implementation of a detailed management plan for the shutdown of the facilities

Optimal shutdown management

International Nuclear Information System (INIS)

Bottasso, C L; Croce, A; Riboldi, C E D

2014-01-01

The paper presents a novel approach for the synthesis of the open-loop pitch profile during emergency shutdowns. The problem is of interest in the design of wind turbines, as such maneuvers often generate design driving loads on some of the machine components. The pitch profile synthesis is formulated as a constrained optimal control problem, solved numerically using a direct single shooting approach. A cost function expressing a compromise between load reduction and rotor overspeed is minimized with respect to the unknown blade pitch profile. Constraints may include a load reduction not-to-exceed the next dominating loads, a not-to-be-exceeded maximum rotor speed, and a maximum achievable blade pitch rate. Cost function and constraints are computed over a possibly large number of operating conditions, defined so as to cover as well as possible the operating situations encountered in the lifetime of the machine. All such conditions are simulated by using a high-fidelity aeroservoelastic model of the wind turbine, ensuring the accuracy of the evaluation of all relevant parameters. The paper demonstrates the capabilities of the novel proposed formulation, by optimizing the pitch profile of a multi-MW wind turbine. Results show that the procedure can reliably identify optimal pitch profiles that reduce design-driving loads, in a fully automated way

Optimal shutdown management

Science.gov (United States)

Bottasso, C. L.; Croce, A.; Riboldi, C. E. D.

2014-06-01

The paper presents a novel approach for the synthesis of the open-loop pitch profile during emergency shutdowns. The problem is of interest in the design of wind turbines, as such maneuvers often generate design driving loads on some of the machine components. The pitch profile synthesis is formulated as a constrained optimal control problem, solved numerically using a direct single shooting approach. A cost function expressing a compromise between load reduction and rotor overspeed is minimized with respect to the unknown blade pitch profile. Constraints may include a load reduction not-to-exceed the next dominating loads, a not-to-be-exceeded maximum rotor speed, and a maximum achievable blade pitch rate. Cost function and constraints are computed over a possibly large number of operating conditions, defined so as to cover as well as possible the operating situations encountered in the lifetime of the machine. All such conditions are simulated by using a high-fidelity aeroservoelastic model of the wind turbine, ensuring the accuracy of the evaluation of all relevant parameters. The paper demonstrates the capabilities of the novel proposed formulation, by optimizing the pitch profile of a multi-MW wind turbine. Results show that the procedure can reliably identify optimal pitch profiles that reduce design-driving loads, in a fully automated way.

Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

International Nuclear Information System (INIS)

Kim, Tae Ryong; Sohn, Seok Man; Lee, Jun Shin; Lee, Sun Ki; Lee, Jong Po

2001-01-01

Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

International Nuclear Information System (INIS)

Goswami, S.

1983-01-01

The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

Shutdown cooling temperature perturbation test for analysis of potential flow blockages

International Nuclear Information System (INIS)

Handbury, J.; Newman, C.; Shynot, T.

1996-01-01

This paper details the methods and results of the 'shutdown cooling test' in October 1995. This novel test was conducted at PLGS while the reactor was shutdown and shutdown cooling (SDC) waster was recirculating to find potential channel blockages resulting from the introduction of wood debris. This test discovered most of the channels that contained major wood and metal debris. (author)

Shutdown Safety in NEK

International Nuclear Information System (INIS)

Gluhak, Mario; Senegovic, Marko

2014-01-01

Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity control, continuity of core cooling, and integrity of fission product barriers are valued as essential, distinguishing attributes of nuclear station work environment'. NEK has recognized the importance of 'defence in depth'Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity

The accidents during shutdown conditions Temelin NPP

International Nuclear Information System (INIS)

Sykora, M.; Mlady, O.

1996-01-01

Two parallel activities oriented for the accidents during shutdown conditions are performed at Temelin NPP: Development of symptom based emergency operating procedures (EOPs) applicable for the accidents which could occur during operational modes 1 through 4; independent evaluation of plant safety as part of the Temelin Shutdown probabilistic assessment to define the accidents which could occur during mode 5 and 6 for which the EOPs must be extended. Both these activities are in progress now because Temelin plant is still in the construction phase

Reactor shutdown device

International Nuclear Information System (INIS)

Matsumiya, Hirohito; Endo, Hiroshi; Tsuboi, Yasushi.

1993-01-01

The present invention concerns a reactor shutdown device capable of suppressing change of a core insertion amount relative to temperature change during normal operation and having a great extension amount due to thermal expansion and high mechanical strength. A control rod main body is contained vertically movably in a guide tube disposed in a reactor core. An extension member extends upward from the upper end of a control rod main body and suspends the control rod main body. A shrinkable member intervenes at a midway of the extension member and is made shrinkable. A temperature sensitive member contains coolants at the inside and surrounds the shrinkable member. Thus, if the temperature of external coolants rises abruptly, the shrinkable member is extended by thermal expansion of the coolants in the temperature sensitive member. Upon usual reactor startup, the coolants in the temperature sensitive member cause no substantial thermal expansion by temperature elevation from a cold shutdown temperature to a rated power operation temperature, and the shrinkable member maintains its original state, so that the control rod main body is not inserted into the reactor core. However, upon abrupt temperature elevation, the control rod main body is inserted into the reactor core. (I.S.)

Industry shutdown rates and permanent layoffs: evidence from firm-worker matched data

Directory of Open Access Journals (Sweden)

Kim P. Huynh

2017-06-01

Full Text Available Abstract Firm shutdown creates a turbulent situation for workers as it leads directly to layoffs for its workers. An additional consideration is whether a firm’s shutdown within an industry creates turbulence for workers at other continuing firms. Using data drawn from the Longitudinal Worker File, a Canadian firm-worker matched employment database, we investigate the impact of industry shutdown rates on workers at continuing firm. This paper exploits variation in shutdown rates across industries and within an industry over time to explain the rate of permanent layoffs and the growth of workers’ earnings. We find an increase in industry shutdown rates increases the probability of permanent layoffs and decreases earnings growth for workers at continuing firms.

Startup and shutdown of the PULSAR Tokamak Reactor

International Nuclear Information System (INIS)

Werley, K.A.; Bathke, C.G.

1994-01-01

Start-up conditions are examined for a pulsed tokamak reactor that uses only inductive plasma current drive for startup, burn and shutdown. A zero-dimensional (profile-averaged) model that describes plasma power and particle balance equations is used to study several aspects of plasma startup and shutdown, including optimization of the startup pathway tradeoff of auxiliary startup heating power versus startup time, volt-second consumtion, thermal stability and partial-power operations

Neutron physical investigations on the shutdown effect of small boronated absorbing spheres for pebble-bed high-temperature gas-cooled reactors

International Nuclear Information System (INIS)

Sgouridis, S.; Schurrer, F.; Muller, H.; Ninaus, W.; Oswald, K.; Neef, R.D.; Schaal, H.

1987-01-01

An emergency shutdown system for high-temperature gas-cooled pebble-bed reactors is proposed in addition to the common absorber rod shutdown system. This system is based on the strongly absorbing effect of small boronated graphite spheres (called KLAK), which trickle in case of emergency by gravity from the top reflector into the reactor core. The inner reflector of the Siemens-Argonaut reactor was substituted by an assembly of spherical Arbeitsgemeinschaft Versuchsreaktor fuel elements, and the shutdown effect was examined by installing well-defined KLAK nests inside this assembly. The purpose was to develop and prove a calculational procedure for determining criticality values for assemblies of large fuel spheres and small absorbing spheres

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.

1987-01-01

The uncertainty in predicting the effectiveness of inherent shutdown (ISD) in innovative designs results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require ISD; (2) the approximation of representing all such ATWS events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. In this summary, the methodology and associated results of work used to establish probabilities of failure of inherent shutdown of innovative LMRs to the unprotected loss-of-flow (LOF) accident are discussed

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

2013-09-30

This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

A Study on Fire Ignition Frequency of UCN 3 during Shutdown

International Nuclear Information System (INIS)

Kim, Kilyoo; Kang, DaeIl; Jang, Seung-Cheol

2014-01-01

A fire ignition frequency of UCN 3 during shutdown, i.e., during POS 3, 4, 5, 6 was calculated by using the new fire PSA method suggested in NUREG/CR-7114. As the fire ignition frequency during full power is calculated by the fixed ignition source and the transient ignition source, the one during shutdown is also calculated by the fixed and the transient ignition source. Since the fixed ignition source was already verified through the walkdown although the walkdown is for the fixed ignition source during full power, additional walkdown for the one during shutdown is not necessary. In the paper, how the fire ignition frequency of UCN 3 during shutdown was calculated is described. A fire ignition frequency of UCN 3 during shutdown, i.e., during POS 3, 4, 5, 6 was calculated by using the new fire PSA method suggested in NUREG/CR-7114. We make the transient ignition fire frequency of each BIN vary according to the daily work order of each POS

40 CFR 63.4168 - What are the requirements for continuous parameter monitoring system installation, operation, and...

Science.gov (United States)

2010-07-01

... divert the emissions away from the add-on control device to the atmosphere. (ii) Car-seal or lock-and-key valve closures. Secure any bypass line valve in the closed position with a car-seal or a lock-and-key... the monitor will indicate valve position. (iv) Automatic shutdown system. Use an automatic shutdown...

Requirements Analysis Study for Master Pump Shutdown System Project Development Specification [SEC 1 and 2

International Nuclear Information System (INIS)

BEVINS, R.R.

2000-01-01

This document has been updated during the definitive design portion of the first phase of the W-314 Project to capture additional software requirements and is planned to be updated during the second phase of the W-314 Project to cover the second phase of the Project's scope. The objective is to provide requirement traceability by recording the analysis/basis for the functional descriptions of the master pump shutdown system. This document identifies the sources of the requirements and/or how these were derived. Each requirement is validated either by quoting the source or an analysis process involving the required functionality, performance characteristics, operations input or engineering judgment

Changing nuclear plant operating limits during startup and shutdown

International Nuclear Information System (INIS)

Arnold, E.C.; Carlson, R.W.; Ray, N.K.; Roarty, D.H.

1990-01-01

During startup and shutdown operation of pressurized water reactor (PWR) nuclear power plants, a low pressure decay heat removal system is used to maintain core cooling. During these phases of operation, there are numerous operating practices and design limits to meet special and sometimes conflicting requirements unique to these operations. This paper evaluates the impact and interdependencies of recent issues on plant operation and design

77 FR 75198 - Standard Format and Content for Post-Shutdown Decommissioning Activities Report

Science.gov (United States)

2012-12-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0299] Standard Format and Content for Post-Shutdown... regulatory guide (DG), DG-1272, ``Standard Format and Content for Post-shutdown Decommissioning Activities... Content for Post-shutdown Decommissioning Activities Report,'' which was issued in July 2000. DG-1271...

40 CFR 63.310 - Requirements for startups, shutdowns, and malfunctions.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 9 2010-07-01 2010-07-01 false Requirements for startups, shutdowns, and malfunctions. 63.310 Section 63.310 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... CATEGORIES National Emission Standards for Coke Oven Batteries § 63.310 Requirements for startups, shutdowns...

76 FR 81998 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2011-12-29

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY..., ``Methodology for Low Power/Shutdown Fire PRA--Draft Report for Comment.'' DATES: Submit comments by March 01... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

Alternative Shutdown Panel. Amaraz Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Saenz de Santa Maria Valin, J.

2016-07-01

Between 2010 and 2014 the Nuclear Power Plant of Almaraz conducted one of the most complex projects in its history: The installation of an Alternative Shutdown Panel with the capability to stop the plant in case of fire in the Control room or in the Cable room. This project represented a great economic and organizational effort for the plant, but at the same time has been a great improvement in the safety of the installation, which was demonstrated by the achievement of a major milestone in the history of Almaraz: The actual shutdown from outside of the Control room. (Author)

Training simulator for advanced gas-cooled reactor (AGR) shutdown sequence equipment

International Nuclear Information System (INIS)

Shankland, J.P.; Nixon, G.L.

1978-01-01

Successful shutdown of nuclear plant is of prime importance for both safety and economic reasons and large sums of money are spent on equipment to make shutdowns fully automatic, thus removing the possibility of operator errors. While this aim can largely be realized, one must consider the possibility of automatic equipment or plant failures when operators are required to take manual action, and off-line training facilities should be available to operating staff to minimize the risk of incorrect actions being taken. This paper presents the practice adopted at Hunterston 'B' Nuclear Power Station to solve this problem and concerns the computer-based training simulator for the Reactor Shutdown Sequence Equipment (RSSE) which was commissioned in January 1977. The plant associated with shutdown is briefly described and the reasoning which shows the need for a simulator is outlined. The paper also gives details of the comprehensive facilities available on the simulator and goes on to describe the form that shutdown training takes and the experience gained at this time. (author)

77 FR 10576 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2012-02-22

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY.../Shutdown Fire PRA.'' In response to request from members of the public, the NRC is extending the public... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused as least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Retrofit of AECL CAN6 seals into the Pickering shutdown cooling pumps

International Nuclear Information System (INIS)

Rhodes, D.; Metcalfe, R.; Brown, G.; Kiameh, P.; Burchett, P.

1997-01-01

The existing mechanical seals in the shutdown cooling (SDC) pumps at the eight-unit Pickering Nuclear Generating Station have caused at least seven forced outages in the last fifteen years. The SDC pumps were originally intended to run only during shutdowns, mostly at low pressure, except for short periods during routine testing of SDC isolation valves while the plant is operating at full pressure to verify that the emergency core injection system is available. Unfortunately, in practice, some SDC pumps must be run much more frequently than this to prevent overheating or freezing of components in the system while the plant is at power. This more severe service has decreased seal lifetime from about 8000 running hours to about 3000 running hours. Rather than tackling the difficult task of eliminating on-power running of the pumps, Pickering decided to install a more robust seal design that could withstand this. Through the process of competitive tender, AECL's CAN6 seal was chosen. This seal has a successful history in similarly demanding conditions in boiling water reactors in the USA. To supplement this and demonstrate there would be no 'surprises,' a 2000-hour test program was conducted. Testing consisted of simulating all the expected conditions, plus some special tests under abnormal conditions. This has given assurance that the seal will operate reliably in the Pickering shutdown cooling pumps. (author)

Quality assurance program plan fuel supply shutdown project

International Nuclear Information System (INIS)

Metcalf, I.L.

1998-01-01

This Quality Assurance Program plan (QAPP) describes how the Fuel Supply Shutdown (FSS) project organization implements the quality assurance requirements of HNF-MP-599, Project Hanford Quality Assurance Program Description (QAPD) and the B and W Hanford Company Quality Assurance Program Plan (QAPP), FSP-MP-004. The QAPP applies to facility structures, systems, and components and to activities (e.g., design, procurement, testing, operations, maintenance, etc.) that could affect structures, systems, and components. This QAPP also provides a roadmap of applicable Project Hanford Policies and Procedures (PHPP) which may be utilized by the FSS project organization to implement the requirements of this QAPP

Concepts in developing technical means of accident shutdown of nuclear reactor

International Nuclear Information System (INIS)

Ionajtis, R.R.; Mikhajlov, M.P.; Cherkashov, Yu.M.

1992-01-01

Logic for realization of multistage (echelon) reactor accident shutdown system (ASS) is proposed on the basis of general safety concepts (OPB-88). ASS includes the basis stage with traditional composition of member systems (executive, control, providing ones), auxiliary (doubling) on the other principle of action and insuring (with direct action). Structural schemes of the system as a whole and member subsystems are presented. Recommendations on developing executive and control subsystems are given

CERN Vacuum-System Activities during the Long Shutdown 1: The LHC’s Injector Chain

CERN Document Server

Ferreira, J A

2014-01-01

During the long shutdown 1 (LS1), several maintenance, consolidation and upgrade activities have been carried out in LHC’s injector chain. Each machine has specific vacuum requirements and different history, which determine the present status of the vacuum components, their maintenance and consolidation needs. The present work presents the priorities agreed at the beginning of the LS1 period and their implementation. Of particular relevance are the interventions in radioactive controlled areas where several leaks due to stress corrosions stopped the operations in the past years. The strategy to reduce the collective dose is presented, in particular the use of remote controlled robots. An important part of the work performed during this period involves supporting other teams (acceptance tests, new equipment installation, etc.). Finally, as a result of the LS1 experience, a medium to long term strategy is depicted, focusing on the preparation of the next shutdown (LS2) and the integration of LINAC4 in the in...

30 CFR 57.8534 - Shutdown or failure of auxiliary fans.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Shutdown or failure of auxiliary fans. 57.8534... Ventilation Underground Only § 57.8534 Shutdown or failure of auxiliary fans. (a) Auxiliary fans installed and... fan maintenance or fan adjustments where air quality is maintained in compliance with the applicable...

Emergency reactor shutdown device

International Nuclear Information System (INIS)

Ikehara, Morihiko.

1982-01-01

Purpose: To smoothen the emergency operation of the control rod in a BWR type reactor and to eliminate the external discharge of radioactively contaminated water. Constitution: A drain receiving tank is connected through a scram valve to the top of a cylinder which is containing a hydraulic piston connected to a trombone-shaped control rod and an accumulator is connected through another scram valve to the bottom of the cylinder. The respective scram valves are constructed to be opened by the reactor emergency shutdown signal from a reactor control system in such a manner that drain valve and a vent valve of the tank normally opened at the standby time are closed after approx. 10 seconds from the opening of the scram valves. In this manner, back pressure is not applied to the hydraulic piston at the emergency time, thereby smoothly operating the control rod. (Sikiya, K.)

Assessment of shutdown management

International Nuclear Information System (INIS)

Marion, A.

1992-01-01

Over the past several years, there has been a number of events that have occurred during nuclear plant outages. These events included losses of AC power, losses of decay heat removal capability, reductions in shutdown margin, and losses of reactor coolant system inventory. Individually, these events have not posed nor indicated an undue risk to public health and safety. Collectively however, they contributed to a perception that outage activities are not being controlled effectively. This paper reports that for many of these same reasons, events that occur during outages have also been of concern to the industry. These events can have a significant economic impact on a company in addition to their being disruptive to the conduct of an efficient outage. And while we have expended industry resources reviewing these events, we have not been fully effective at addressing the root cause of the problem

PSA for the shutdown mode for nuclear power plants

International Nuclear Information System (INIS)

1994-06-01

The meeting, which was attended by more than 75 participants from 20 countries, provided a broad discussion forum where all the currently active major shutdown PSA programmes were reviewed. The meeting also addressed the issues related to actual performance of shutdown PSA studies as well as insight gained from the studies. This document, which was prepared during the TCM, contains the results of extensive discussions which were held in specific working groups. The papers presented at the meeting provide a comprehensive overview of the state of the art of shutdown risk assessment and remedial measures taken to reduce the risk in outages. It is hoped that this document will be very useful to all individuals with interest in increasing safety during outages at NPPs. Refs, figs and tabs

Core shutdown report: Subcycle K-14.1

International Nuclear Information System (INIS)

Gough, S.T.

1992-05-01

When a reactor is shut down, there is a set of rules that must be followed to guarantee that the reactor remains in a safe shutdown state. Some of these rules involve the cooling of heat generating assemblies before, during, and after charge-discharge (C ampersand D) operations. These rules ensure that C ampersand D operations will not endanger the integrity of the fuel or targets by allowing them to overheat. DPSOL 105-1225, Assembly Discharge and Forced Cooling Requirements, is the primary operations procedure that governs these cooling rules. The specific shutdown cooling limits that are input into this procedure are contained within this report

Mobility problems at the KNK II shut-down systems, cause investigations and valuation in comparison with the experience at other plants

International Nuclear Information System (INIS)

Hess, B.

1992-12-01

During the operation of the second core of the fast test reactor KNK II the shutdown systems showed repeatedly problems with their mobility, which also caused to be reported events. The present report gives a summary description of the events in chronological order. The investigations to remove the mobility problems and the resulting design modifications are described together with the comments of the licensing authorities on the way to the restart of the plant. The results of the post-irradiation investigations in the hot cells and of sodium-chemical investigations are also described. In addition to the comparison of the events at the KNK plant itself and a review of the experiences at comparable plants it will be shown that all known cases of mobility problems did only influence the availability of the plant but that the safe shut-down of the plant was never at risk [de

Design of emergency shutdown system for the Tehran Research Reactor; Part I: Neutronics investigation

International Nuclear Information System (INIS)

Safarinia, M.; Faghihi, F.; Mirvakili, S.M.; Fakhraei, A.

2017-01-01

Highlights: • An emergency shutdown system for the TRR is carried out based on a heavy water tank. • The performance of the heavy water tank are carried out based on “first and equilibrium cores”. • Heavy water discharging flow rate is also studied in the current research. • Thermal flux in the radioisotope channel with and without the heavy water tank are studied. • A core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). - Abstract: In this paper, a neutronics design of the secondary (i.e., emergency) shutdown system for the Tehran Research Reactor (TRR) is carried out based on a heavy water tank design. The heavy water tank in a cylindrical shape is around the core, and calculations for the optimized radius and height of the tank are performed. The performance of the heavy water tank calculations are carried out based on two types of fuel loading, which are called the “first and equilibrium cores” of the TRR. For both cases, neutronics and standard safety analysis are taken into account, benchmarked, and described herein. Heavy water discharging flow rate is also studied in the current research, and the results are compared with the IAEA criteria. Moreover, thermal flux in the radioisotope channel with and without the heavy water tank (as the reflector) are studied herein. Specifically, a core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). Based on our optimization, the 5 × 5 fuel assembly, which is called “B configuration,” has better performance and efficiency than that of the other described layouts.

Analysis of failure dependent test, repair and shutdown strategies for redundant trains

International Nuclear Information System (INIS)

Uryasev, S.; Samanta, P.

1994-09-01

Failure-dependent testing implies a test of a redundant components (or trains) when failure of one component has been detected. The purpose of such testing is to detect any common cause failures (CCFs) of multiple components so that a corrective action such as repair or plant shutdown can be taken to reduce the residence time of multiple failures, given a failure has been detected. This type of testing focuses on reducing the conditional risk of CCFs. Formulas for calculating the conditional failure probability of a two train system with different test, repair and shutdown strategies are developed. A methodology is presented with an example calculation showing the risk-effectiveness of failure-dependent strategies for emergency diesel generators (EDGs) in nuclear power plants (NPPs)

Development of Abnormal Operating Strategies for Station Blackout in Shutdown Operating Mode in Pressurized Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hwang, Su-Hyun [FNC Tech. Co., Yongin (Korea, Republic of)

2016-10-15

Loss of all AC power is classified as one of multiple failure accident by regulatory guide of Korean accident management program. Therefore we need develop strategies for the abnormal operating procedure both of power operating and shutdown mode. This paper developed abnormal operating guideline for loss of all AC power by analysis of accident scenario in pressurized water reactor. This paper analyzed the loss of ultimate heat sink (LOUHS) in shutdown operating mode and developed the operating strategy of the abnormal procedure. Also we performed the analysis of limiting scenarios that operator actions are not taken in shutdown LOUHS. Therefore, we verified the plant behavior and decided operator action to taken in time in order to protect the fuel of core with safety. From the analysis results of LOUHS, the fuel of core maintained without core uncovery for 73 minutes respectively for opened RCS states after the SBO occurred. Therefore, operator action for the emergency are required to take in 73 minutes for opened RCS state. Strategy is to cooldown by using spent fuel pool cooling system. This method required to change the plant design in some plant. In RCS boundary closed state, first abnormal operating strategy in shutdown LOUHS is first abnormal operating strategy in shutdown LOUHS is to remove the residual heat of core by steam dump flow and auxiliary feedwater of SG.

The Alternative Design Features for Safety Enhancement in Shutdown Operation

International Nuclear Information System (INIS)

Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro

2009-01-01

PSA can be used to confirm that the new plant design is complied with the applicable safety goals, and to select among the alternate design options. A shutdown PSA provides insight for outage planning schedule, outage management practices, and design modifications. Considering the results of both LPSD PSA studies and operating experiences for low power and shutdown, the improvements can be proposed to reduce the high risk contribution. The improvements/enhancements during shutdown operation may be divided into categories such as hardware, administrative management, and operational procedure. This paper presents on an example how the risk related to an accidental situation can be reduced, focusing the hardware design changes for the newly designed NPPs

Study of methodology for low power/shutdown fire PSA

International Nuclear Information System (INIS)

Yan Zhen; Li Zhaohua; Li Lin; Song Lei

2014-01-01

As a risk assessment technology based on probability, the fire PSA is accepted abroad by nuclear industry in its application in the risk assessment for nuclear power plants. Based on the industry experience, the fire-induced impact on the plant safety during low power and shutdown operation cannot be neglected, therefore fire PSA can be used to assess the corresponding fire risk. However, there is no corresponding domestic guidance/standard as well as accepted analysis methodology up to date. Through investigating the latest evolvement on fire PSA during low power and shutdown operation, and integrating its characteristic with the corresponding engineering experience, an engineering methodology to evaluate the fire risk during low power and shutdown operation for nuclear power plant is established in this paper. In addition, an analysis demonstration as an example is given. (authors)

Passive shut-down of ITER plasma by Be evaporation

International Nuclear Information System (INIS)

Amano, Tsuneo.

1996-02-01

In an accident event where the cooling system of first wall of the ITER fails, the first wall temperature continues to rise as long as the ignited state of the core plasma persists. In this paper, a passive shut-down scheme of the ITER from this accident by evaporated Be from the first wall is examined. It is shown the estimated Be influx 5 10 24 /sec is sufficient to quench the ignition. (author)

Fuse and application of said fuse to the construction of an emergency shutdown system for a nuclear reactor

International Nuclear Information System (INIS)

Taulier, H.H.L.; Brugeille, G.

1978-01-01

A fuse device for an automatic emergency shutdown system in fast reactors provides a coupling between a casing tube placed within a fuel can and a series of neutron-absorbing masses held together above the reactor core under normal operating conditions but released in free fall to the lower portion of the casing tube at the level of the reactor core as a result of melting of the fuse when operating characteristics such as temperature or neutron flux attain a level which exceeds a predetermined threshold

Fuse and application of said fuse to the construction of an emergency shutdown system for a nuclear reactor

International Nuclear Information System (INIS)

Taulier, H.H.L.; Brugeilles, G.

1976-01-01

A fuse device for an automatic emergency shutdown system in fast reactors provides a coupling between a casing tube placed within a fuel can and a series of neutron-absorbing masses held together above the reactor core under normal operating conditions. They are released in free fall to the lower portion of the casing tube at the level of the reactor core as a result of melting of the fuse when operating characteristics such as temperature or neutron flux attain a level which exceeds a predetermined threshold

Study on the Post-Fire Safe-Shutdown Analysis for CANDU NPPs

Energy Technology Data Exchange (ETDEWEB)

Kim, In Hwan; Kim, Yun Jung; Park, Mun Hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The purpose of this paper is to study a method of the Post-Fire Safe-Shutdown Analysis in order to apply to CANDU NPPs when one group of the Safety Structures, Systems and Components(SCCs) is failed by Fire. The purpose of Fire Protection is prevention, suppression of the fire and mitigation of the effect on the Nuclear Safety. When fire takes place at the Nuclear Power Plants(NPPs), the reactor should achieve and maintain safe shut-down condition and minimize radioactive material release to an environment. The purpose of the Post-Fire SSA process is an evaluation process during a fire at NPPs. At this study, the process was conceptually adopted for control room complex of CANDU NPPs. The Core Damage Frequency of the Reactor will be evaluated more accurately if the SSA is adopted adequately at a fire.

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C Systems

Energy Technology Data Exchange (ETDEWEB)

Jang, Seung Cheol; Kang, Hyung Gook; Lim, Ho Gon; Park, Jin Hee; Eom, Heung Sub; Kim, Tae Woon; Ha, Jae Joo

2005-04-15

There are two technical areas to deal with in the project; the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: Quality assessment of a LPSD PSA model for a Korean Standard Nuclear Power Plant (KSNP), Quality improvement of the KSNP LPSD PSA model in the following four technical areas; plant operating status (POS), initiating event analysis, determination of success criteria, accident sequence analysis, Development of the LPSD risk management technologies, Unavailability analysis of Digital safety systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS), Impact analysis of the digital safety systems on plant risks throughout of the digital plant risk models for evaluating core damage frequency (CDF) and large early release frequency (LERF), Study on the methodologies for treating digital-specific problems in the digital I and C PSA such as reliability of safety-critical software, common cause failure (CCF) of digital components, fault coverage, etc.

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C Systems

International Nuclear Information System (INIS)

Jang, Seung Cheol; Kang, Hyung Gook; Lim, Ho Gon; Park, Jin Hee; Eom, Heung Sub; Kim, Tae Woon; Ha, Jae Joo

2005-04-01

There are two technical areas to deal with in the project; the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: Quality assessment of a LPSD PSA model for a Korean Standard Nuclear Power Plant (KSNP), Quality improvement of the KSNP LPSD PSA model in the following four technical areas; plant operating status (POS), initiating event analysis, determination of success criteria, accident sequence analysis, Development of the LPSD risk management technologies, Unavailability analysis of Digital safety systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS), Impact analysis of the digital safety systems on plant risks throughout of the digital plant risk models for evaluating core damage frequency (CDF) and large early release frequency (LERF), Study on the methodologies for treating digital-specific problems in the digital I and C PSA such as reliability of safety-critical software, common cause failure (CCF) of digital components, fault coverage, etc

Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

International Nuclear Information System (INIS)

Kim, I.S.; Samanta, P.K.

1994-01-01

When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require ''immediate'' shutdown of the plant. In this paper, the authors present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion

Improving the action requirements of technical specifications: A risk-comparison of continued operation and plant shutdown

Energy Technology Data Exchange (ETDEWEB)

Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States); Mankamo, T.

1995-04-01

When the systems needed to remove decay heat are inoperable or degraded, the risk of shutting down the plant may be comparable to, or even higher than, that of continuing power operation with the equipment inoperable while giving priority to repairs. This concern arises because the plant may not have sufficient capability for removing decay heat during the shutdown. However, Technical Specifications (TSs) often require {open_quotes}immediate{close_quotes} shutdown of the plant. In this paper, we present risk-based analyses of the various operational policy alternatives available in such situations, with an example application to the standby service water (SSW) system of a BWR. These analyses can be used to define risk-effective requirements for those standby safety systems under discussion.

Risk impact of BWR technical specifications requirements during shutdown

International Nuclear Information System (INIS)

Staple, B.D.; Kirk, H.K.; Yakle, J.

1994-10-01

This report presents an application of probabilistic models and risk based criteria for determining the risk impact of the Limiting Conditions of Operations (LCOs) in the Technical Specifications (TSs) of a boiling water reactor during shutdown. This analysis studied the risk impact of the current requirements of Allowed Outage Times (AOTs) and Surveillance Test Intervals (STIs) in eight Plant Operational States (POSs) which encompass power operations, shutdown, and refueling. This report also discusses insights concerning TS action statements

Loss-of-benefits analysis for nuclear power plant shutdowns: methodology and illustrative case study

International Nuclear Information System (INIS)

Peerenboom, J.P.; Buehring, W.A.; Guziel, K.A.

1983-11-01

A framework for loss-of-benefits analysis and a taxomony for identifying and categorizing the effects of nuclear power plant shutdowns or accidents are presented. The framework consists of three fundamental steps: (1) characterizing the shutdown; (2) identifying benefits lost as a result of the shutdown; and (3) quantifying effects. A decision analysis approach to regulatory decision making is presented that explicitly considers the loss of benefits. A case study of a hypothetical reactor shutdown illustrates one key loss of benefits: net replacement energy costs (i.e., change in production costs). Sensitivity studies investigate the responsiveness of case study results to changes in nuclear capacity factor, load growth, fuel price escalation, and discount rate. The effects of multiple reactor shutdowns on production costs are also described

40 CFR 65.6 - Startup, shutdown, and malfunction plan and procedures.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Startup, shutdown, and malfunction... (CONTINUED) AIR PROGRAMS (CONTINUED) CONSOLIDATED FEDERAL AIR RULE General Provisions § 65.6 Startup... Group 2A or Group 2B process vents. (b) Startup, shutdown, and malfunction plan—(1) Description and...

Oak Ridge Research Reactor shutdown maintenance and surveillance

International Nuclear Information System (INIS)

Coleman, G.H.; Laughlin, D.L.

1990-10-01

The Department of Energy ordered the Oak Ridge Research Center Reactor to be placed in permanent shutdown on July 14, 1987. Maintenance activities, both mechanical and instrument, were essentially routine in nature. The performance of the instrumentation for the facility was satisfactory, and maintenance required is provided. The performance of the process system was satisfactory, and maintenance required is indicated. The results of efficiency tests of the various gaseous-waste filters have been summarized and preparations for transfer of the facility to the remedial action program is also indicated

Loss of benefits resulting from mandated nuclear plant shutdowns

International Nuclear Information System (INIS)

Peerenboom, J.P.; Buehring, W.A.

1982-01-01

This paper identifies and discusses some of the important consequences of nuclear power plant unavailability, and quantifies a number of technical measures of loss of benefits that result from regulatory actions such as licensing delays and mandated nuclear plant outages. The loss of benefits that accompany such regulatory actions include increased costs of systems generation, increased demand for nonnuclear and often scarce fuels, and reduced system reliability. This paper is based on a series of case studies, supplemented by sensitivity studies, on hypothetical nuclear plant shutdowns. These studies were developed by Argonne in cooperation with four electric utilities

Reliability of Offshore Wind Turbine Drivetrains based on Measured Shut-down Events

DEFF Research Database (Denmark)

Natarajan, Anand; Buhl, Thomas

2015-01-01

by initiating blade pitching to feather and also sometimes using the generator torqueas a brake mechanism. The shutdowns due to wind speed variation nearcut-out are predicted using an Inverse First Order Reliability Model(IFORM) whereby an expected annual frequency of normal shutdownsat cut-out is put forth...... normal operation and with shutdowns. The maximum coefficient of variation (CoV) due to varying wind conditions was found on the low speed shaft torsion, but the shutdowns by themselves were not seento significantly change the fatigue loads....

40 CFR 62.15150 - What happens to the operating requirements during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... requirements during periods of startup, shutdown, and malfunction? 62.15150 Section 62.15150 Protection of... § 62.15150 What happens to the operating requirements during periods of startup, shutdown, and... municipal waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or...

A Fast Shutdown Technique for Large Tokamaks

International Nuclear Information System (INIS)

Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.

1999-01-01

A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR

Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

International Nuclear Information System (INIS)

Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

2009-01-01

The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

40 CFR 60.1695 - What happens to the operating requirements during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... requirements during periods of startup, shutdown, and malfunction? 60.1695 Section 60.1695 Protection of... Requirements § 60.1695 What happens to the operating requirements during periods of startup, shutdown, and... municipal waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or...

Maintenance of shutdown system in the reactor core to minimize the radioactive waste generation

International Nuclear Information System (INIS)

Ponzoni Filho, P.; Fernandes, V.B.

1988-01-01

This paper recommends a modification on the actual strategy of going from Cold-Shutdown to Critical, that will save about 6000 liter of boric acid and 30,000 liters of demineralized water for each reactor criticalization. This strategy will reduce the radioactive waste disposal volume to only about 5% of what would be generated following the actual strategy. (author) [pt

77 FR 73968 - Reconsideration of Certain New Source and Startup/Shutdown Issues: National Emission Standards...

Science.gov (United States)

2012-12-12

...; FRL-9762-1] RIN 2060-AR62 Reconsideration of Certain New Source and Startup/Shutdown Issues: National... Source and Startup/Shutdown Issues: National Emission Standards for Hazardous Air Pollutants from Coal... November 30, 2012, proposed ``Reconsideration of Certain New Source and Startup/Shutdown Issues: National...

40 CFR 60.1220 - What happens to the emission limits during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... during periods of startup, shutdown, and malfunction? 60.1220 Section 60.1220 Protection of Environment... Emission Limits § 60.1220 What happens to the emission limits during periods of startup, shutdown, and... waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or malfunction must...

Technical Specification action statements requiring shutdown

International Nuclear Information System (INIS)

Mankamo, T.; Kim, I.S.; Samanta, P.K.

1993-11-01

When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements

SEPRA - shutdown PSA for the OLKILUOTO nuclear power plant

International Nuclear Information System (INIS)

Himanen, R.

1995-01-01

The utility TVO has extended the PSA study to the analysis of refueling, shutdown and startup. The Shutdown Event PRA (SEPRA) was reported to the authority in September 1992. The study consists of the analysis of leaks and loss of decay heat removal in the planned shutdown conditions. Special studies were performed for the cold pressurization, for local criticality events, for heavy load transport and for the transients during startup and shutdown. A remarkable effort was put to identify risks, i.e. to the qualitative analysis. The regular preventive maintenance tasks in the refueling outages were analyzed and the important tasks were selected for further studies. Besides the severe core damage risk the utility was interested in less grave consequences, e.g. the economic risks, causing significant extension of outages. The plant specific screening of initiators consisted of a study on the incident history and of interviewing the plant personnel on selected tasks. A number of thermohydraulic calculations were carried out to support the analysis of accident sequences. The operator actions after an initiating event were verified with the operating staff. The annual core damage risk from the refueling outage is about one forth of the total annual risk. The modifications decreased significantly the core damage frequency. It is foreseen that the SEPRA will form a basis of the procedure enhancement for the low power states. (author) 5 figs., 1 tab., 10 refs

Correlation of operating parameters on turbine shaft vibrations

Science.gov (United States)

Dixit, Harsh Kumar; Rajora, Rajeev

2016-05-01

The new generation of condition monitoring and diagnostics system plays an important role in efficient functioning of power plants. In most of the rotating machine, defects can be detected by such a system much before dangerous situation occurs. It allows the efficient use of stationary on-line continuous monitoring system for condition monitoring and diagnostics as well. Condition monitoring of turbine shaft can not only reduce expenses of maintenance of turbo generator of power plants but also prevents likely shutdown of plant, thereby increases plant load factor. Turbo visionary parameters are essential part of health diagnosis system of turbo generator. Particularly steam pressure, steam temperature and lube oil temperature are important parameters to monitor because they are having much influence on turbine shaft vibration and also governing systems are available for change values of those parameters. This paper includes influence of turbo visionary parameters i.e., steam temperature, steam pressure, lube oil temperature, turbine speed and load on turbine shaft vibration at turbo generator at 195 MW unit-6,Kota Super Thermal Power Station by measuring vibration amplitude and analyze them in MATLAB.

Evolution of the ATLAS distributed computing system during the LHC long shutdown

Science.gov (United States)

Campana, S.; Atlas Collaboration

2014-06-01

The ATLAS Distributed Computing project (ADC) was established in 2007 to develop and operate a framework, following the ATLAS computing model, to enable data storage, processing and bookkeeping on top of the Worldwide LHC Computing Grid (WLCG) distributed infrastructure. ADC development has always been driven by operations and this contributed to its success. The system has fulfilled the demanding requirements of ATLAS, daily consolidating worldwide up to 1 PB of data and running more than 1.5 million payloads distributed globally, supporting almost one thousand concurrent distributed analysis users. Comprehensive automation and monitoring minimized the operational manpower required. The flexibility of the system to adjust to operational needs has been important to the success of the ATLAS physics program. The LHC shutdown in 2013-2015 affords an opportunity to improve the system in light of operational experience and scale it to cope with the demanding requirements of 2015 and beyond, most notably a much higher trigger rate and event pileup. We will describe the evolution of the ADC software foreseen during this period. This includes consolidating the existing Production and Distributed Analysis framework (PanDA) and ATLAS Grid Information System (AGIS), together with the development and commissioning of next generation systems for distributed data management (DDM/Rucio) and production (Prodsys-2). We will explain how new technologies such as Cloud Computing and NoSQL databases, which ATLAS investigated as R&D projects in past years, will be integrated in production. Finally, we will describe more fundamental developments such as breaking job-to-data locality by exploiting storage federations and caches, and event level (rather than file or dataset level) workload engines.

Evolution of the ATLAS distributed computing system during the LHC long shutdown

International Nuclear Information System (INIS)

Campana, S

2014-01-01

The ATLAS Distributed Computing project (ADC) was established in 2007 to develop and operate a framework, following the ATLAS computing model, to enable data storage, processing and bookkeeping on top of the Worldwide LHC Computing Grid (WLCG) distributed infrastructure. ADC development has always been driven by operations and this contributed to its success. The system has fulfilled the demanding requirements of ATLAS, daily consolidating worldwide up to 1 PB of data and running more than 1.5 million payloads distributed globally, supporting almost one thousand concurrent distributed analysis users. Comprehensive automation and monitoring minimized the operational manpower required. The flexibility of the system to adjust to operational needs has been important to the success of the ATLAS physics program. The LHC shutdown in 2013-2015 affords an opportunity to improve the system in light of operational experience and scale it to cope with the demanding requirements of 2015 and beyond, most notably a much higher trigger rate and event pileup. We will describe the evolution of the ADC software foreseen during this period. This includes consolidating the existing Production and Distributed Analysis framework (PanDA) and ATLAS Grid Information System (AGIS), together with the development and commissioning of next generation systems for distributed data management (DDM/Rucio) and production (Prodsys-2). We will explain how new technologies such as Cloud Computing and NoSQL databases, which ATLAS investigated as R and D projects in past years, will be integrated in production. Finally, we will describe more fundamental developments such as breaking job-to-data locality by exploiting storage federations and caches, and event level (rather than file or dataset level) workload engines.

The Bulgaria before shut-down of next two blocks

International Nuclear Information System (INIS)

Dobak, D.

2005-01-01

The Ministry of Trade and Industry of United Kingdom in the frame of realization of programmes for the Middle and East Europe in the area of nuclear energetics during October 5 - 7, 2005 in Kozloduj has organized the Second International Conference on the theme 'Liquidation, social and economic changes'. In this paper author informs about Kozloduj NPP and plans for shut-down of this NPP as well as consequences of the shut-down. One of them the increase of unemployment and social impact for this region are presented

Electricity-market price and nuclear power plant shutdown: Evidence from California

International Nuclear Information System (INIS)

Woo, C.K.; Ho, T.; Zarnikau, J.; Olson, A.; Jones, R.; Chait, M.; Horowitz, I.; Wang, J.

2014-01-01

Japan's Fukushima nuclear disaster, triggered by the March 11, 2011 earthquake, has led to calls for shutting down existing nuclear plants. To maintain resource adequacy for a grid's reliable operation, one option is to expand conventional generation, whose marginal unit is typically fueled by natural-gas. Two timely and relevant questions thus arise for a deregulated wholesale electricity market: (1) what is the likely price increase due to a nuclear plant shutdown? and (2) what can be done to mitigate the price increase? To answer these questions, we perform a regression analysis of a large sample of hourly real-time electricity-market price data from the California Independent System Operator (CAISO) for the 33-month sample period of April 2010–December 2012. Our analysis indicates that the 2013 shutdown of the state's San Onofre plant raised the CAISO real-time hourly market prices by $6/MWH to $9/MWH, and that the price increases could have been offset by a combination of demand reduction, increasing solar generation, and increasing wind generation. - Highlights: • Japan's disaster led to calls for shutting down existing nuclear plants. • We perform a regression analysis of California's real-time electricity-market prices. • We estimate that the San Onofre plant shutdown has raised the market prices by $6/MWH to $9/MWH. • The price increases could be offset by demand reduction and renewable generation increase

Application of PSA to reduce frequency of unplanned shutdown of the reactor

International Nuclear Information System (INIS)

Tanipanichskul, P.

1988-08-01

The relative importance of all the operating and safety systems of the reactor TRR-1/M1 as well as the major failure modes of the systems are pointed out. The average unavailability of the reactor is 3·3 E-2 per cycle of operation which is in the range value of the actual reactor shutdown recorded during normal operation. Some guidance for annual maintenance and also suggestions for system development to increase safety systems reliability are determined. PSA was applied to improve the safety systems reliability of an operating research reactor. Refs, tabs

Standardization of the time for the execution of HANARO start-up and shutdown procedures

International Nuclear Information System (INIS)

Choi, H. Y.; Lim, I. C.; Hwang, S. R.; Kang, T. J.; Youn, D. B.

2003-01-01

For the standardization of the time to execute HANARO start-up and shutdown procedures, code names were assigned to the individual procedures and the work time were investigated. The data recorded by the operators during start-up and shutdown were statistically analyzed. The analysis results will be used for the standardization of start-up and shutdown procedures and it will be reflected in the procedure document

Probabilistic safety assessments of nuclear power plants for low power and shutdown modes

International Nuclear Information System (INIS)

2000-03-01

Within the past several years the results of nuclear power plant operating experience and performance of probabilistic safety assessments (PSAs) for low power and shutdown operating modes have revealed that the risk from operating modes other than full power may contribute significantly to the overall risk from plant operations. These early results have led to an increased focus on safety during low power and shutdown operating modes and to an increased interest of many plant operators in performing shutdown and low power PSAs. This publication was developed to provide guidance and insights on the performance of PSA for shutdown and low power operating modes. The preparation of this publication was initiated in 1994. Two technical consultants meetings were conducted in 1994 and one in February 1999 in support of the development of this report

The shutdown reactor: Optimizing spent fuel storage cost

International Nuclear Information System (INIS)

Pennington, C.W.

1995-01-01

Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wet and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec's findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest

Management of accidental scenarios involving the loss of RHRS under shutdown conditions

International Nuclear Information System (INIS)

Serradell, V.; Villanueva, J.F.; Martorell, S.; Carlos, S.; Pelayo, F.; Mendizabal, R.; Sol, I.

2009-01-01

Results from current Probabilistic Safety Assessment studies of Nuclear Power Plants show the importance of some risky scenarios with the plant at low power and shutdown conditions as compared to the accident scenarios with the plant operating at full power. Technical Specifications establish the Limiting Conditions for operation to assure the plant integrity in each Plant Operational State (POS). Moreover, the plant configuration may differ from the beginning to the end of a certain Plant Operational State, so the Limiting Conditions for Operation (LCO) established could be revised as, depending on the plant configuration, the transient evolution may be slightly different. For a PWR plant, one of the most risky accidental sequences in shutdown is the loss of the residual heat removal system, Using the information provided by the plant low power probabilistic safety analysis (LPSA), which should address the Limiting Conditions for Operation imposed by the current Technical Specification, two situations are distinguished: Main Reactor Cooling System (RCS) fully filled with water and RCS partially filled. In addition, while the primary system is partially filled in Cold Shutdown, two different plant configurations can be distinguished, which depend on the particular POS: RCS open and closed. For each case, the corresponding Technical Specification establishes the path to evacuate the residual heat generated. This paper explores the possibility of having alternative or complementary sources for heat removal others than the ones established in the Technical Specification. Especial attention is paid to the role of Steam Generators as an effective heat sink and the possibility of restart of the redundant RHR train. Such alternatives will influence LPSA implementation results. To perform this analysis the loss of the RHR system in a PWR plant has been simulated using RELAP-5 considering the plant in different plant operational states. One of the main results of this work

MCR2S unstructured mesh capabilities for use in shutdown dose rate analysis

International Nuclear Information System (INIS)

Eade, T.; Stonell, D.; Turner, A.

2015-01-01

Highlights: • Advancements in shutdown dose rate calculations will be needed as fusion moves from experimental reactors to full scale demonstration reactors in order to ensure the safety of personnel. • The MCR2S shutdown dose rate tool has been modified to allow shutdown dose rates calculations using an unstructured mesh. • The unstructured mesh capability of MCR2S was used on three shutdown dose rate models, a simple sphere, the ITER computational benchmark and the DEMO computational benchmark. • The results showed a reasonable agreement between an unstructured mesh approach and the CSG approach and highlighted the need to carefully choose the unstructured mesh resolution. - Abstract: As nuclear fusion progresses towards a sustainable energy source and the power of tokamak devices increases, a greater understanding of the radiation fields will be required. As well as on-load radiation fields, off-load or shutdown radiation field are an important consideration for the safety and economic viability of a commercial fusion reactor. Previously codes such as MCR2S have been written in order to predict the shutdown dose rates within, and in regions surrounding, a fusion reactor. MCR2S utilises a constructive solid geometry (CSG) model and a superimposed structured mesh to calculate 3-D maps of the shutdown dose rate. A new approach to MCR2S calculations is proposed and implemented using a single unstructured mesh to replace both the CSG model and the superimposed structured mesh. This new MCR2S approach has been demonstrated on three models of increasing complexity. These models were: a sphere, the ITER computational shutdown dose rate benchmark and the DEMO computational shutdown dose rate benchmark. In each case the results were compared to MCR2S calculations performed using MCR2S with CSG geometry and a superimposed structured mesh. It was concluded that the results from the unstructured mesh implementation of MCR2S compared well to the CSG structured mesh

78 FR 38739 - Standard Format and Content for Post-Shutdown Decommissioning Activities Report

Science.gov (United States)

2013-06-27

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0299] Standard Format and Content for Post-Shutdown Decommissioning Activities Report AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance..., ``Standard Format and Content for Post-shutdown Decommissioning Activities Report.'' This guide describes a...

Development of Start-up and Shutdown Procedure for the HANARO Fuel Test Loop

International Nuclear Information System (INIS)

Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, J. M.; Lee, C. Y.; Ahn, S. H.

2009-06-01

A start-up and shutdown procedure for the HANARO fuel test loop has been developed. This is a facility for fuel and material irradiation tests. The facility provides experimental conditions similar to the normal operational pressures and temperatures of commercial PWR and CANDU plants. The normal operation modes of the HANARO fuel test loop are classified into loop shutdown, cold stand-by 1, cold stand-by 2, hot stand-by, and hot operation. The operation modes depend on the fission power of test fuels and the coolant temperature at the inlet of the in-pile test section. The HANARO must maintain a shutdown mode if the HANARO fuel test loop is loop shutdown, cold stand-by 1, cold stand-by 2, or hot stand-by. As the HANARO becomes power operation mode, the operation mode of the HANARO fuel test loop comes to hot operation from hot stand-by. The procedure for the HANARO fuel test loop consists of four main parts such as check of initial conditions, stat-up operation procedure, shutdown operation procedure, and check lists for operations. Several hot test operations ensure that the procedure is appropriate

Analysis of HFETR shut-down state caused by loss of off-site power supply

International Nuclear Information System (INIS)

Wang Jinghu

1997-01-01

During the last 15 years, there are more than 40 unplanned shut-downs caused by loss of off-site power in HFETR. Because HFETR is a special research reactor, the author describes the shut-down state as three period. The author also discusses the influence of the number of shut-down due to loss of off-site power supply on the reactor safety, and propose some suggestions and measures to reduce the effects

Global shutdown dose rate maps for a DEMO conceptual design

International Nuclear Information System (INIS)

Leichtle, D.; Pereslavtsev, P.; Sanz, J.; Catalan, J.P.; Juarez, R.

2015-01-01

Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

Global shutdown dose rate maps for a DEMO conceptual design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pereslavtsev, P. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sanz, J.; Catalan, J.P.; Juarez, R. [Universidad Nacional de Educación a Distancia(UNED), E.T.S. Ingenieros Industriales, C/ Juan del Rosal 12, 28040 Madrid (Spain)

2015-10-15

Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

CV activities on the LHC complex during the long shutdown

CERN Document Server

Deleval, S; Body, Y; Obrecht, M; Moccia, S; Peon, G

2011-01-01

The presentation gives an overview of the major projects and work foreseen to be performed during next long shutdown on cooling and ventilation plants. Several projects are needed following the experience of the last years when LHC was running, in particular the modifications in the water cooling circuits presently in overflow. Some other projects are linked to the CV consolidation plan. Finally, most of the work shall be done to respond to additional requests: SR buildings air conditioning, the need to be able to clean and maintain the LHC cooling towers without a complete stop of cooling circuits, the upgrade of the air conditioning of the CCC rack room cooling etc. For all these activities, the author will detail constraints and the impact on the schedule and on the operation of the plants that will however need to run for most of the shutdown duration. The consequence of postponing the long shutdown from 2012 to 2013 will be also covered.

40 CFR 63.2852 - What is a startup, shutdown, and malfunction plan?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 12 2010-07-01 2010-07-01 true What is a startup, shutdown, and... Production Compliance Requirements § 63.2852 What is a startup, shutdown, and malfunction plan? You must...)(2) malfunction period, or the § 63.2850(c)(2) or (d)(2) initial startup period. The SSM plan must...

40 CFR 60.2918 - What happens during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 6 2010-07-01 2010-07-01 false What happens during periods of startup... of startup, shutdown, and malfunction? The emission limitations and operating limits apply at all times except during OSWI unit startups, shutdowns, or malfunctions. Performance Testing ...

Calculation of the negative reactivity inserted by the shutdown system number two (SDS2) of a CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Arsenault, B [Ecole Polytechnique, Montreal, PQ (Canada)

1994-12-31

The secondary shutdown system (SDS2) of a CANDU reactor consists of liquid poison injection through nozzles disposed horizontally across the core. The nominal concentration of gadolinium nitrate poison is 8000 ppm. With the methods available to the nuclear industry for calculating the negative reactivity inserted by the SDS2, some approximations are needed, and a simplified model of poison propagation has to be used to calculate the differential cross sections. The objective of this paper is to evaluate the errors introduced by the approximations in the supercell and core calculations. The MULTICELL and EXCELL codes gave different power distributions, and further work was recommended. 9 refs., 2 tabs., 4 figs.

Using dew points to estimate savings during a planned cooling shutdown

Science.gov (United States)

Friedlein, Matthew T.; Changnon, David; Musselman, Eric; Zielinski, Jeff

2005-12-01

In an effort to save money during the summer of 2003, Northern Illinois University (NIU) administrators instituted a four-day working week and stopped air conditioning buildings for the three-day weekends (Friday through Sunday). Shutting down the air conditioning systems caused a noticeable drop in electricity usage for that part of the campus that features in our study, with estimated total electricity savings of 1,268,492 kilowatt-hours or 17% of the average usage during that eight-week period. NIU's air conditioning systems, which relied on evaporative cooling to function, were sensitive to dew point levels. Greatest savings during the shutdown period occurred on days with higher dew points. An examination of the regional dew point climatology (1959 2003) indicated that the average summer daily dew point for 2003 was 14.9°C (58.8°F), which fell in the lowest 20% of the distribution. Based on the relationship between daily average dew points and electrical usage, a predictive model that could estimate electrical daily savings was created. This model suggests that electrical savings related to any future three-day shutdowns over summer could be much greater in more humid summers. Studies like this demonstrate the potential value of applying climatological information and of integrating this information into practical decision-making.

The application and design of distributed control system in reactor shutdown system of Qinshan phase III

International Nuclear Information System (INIS)

Su Guoquan; Liu Wangtian; Yu Yijun; Xiong Weihua

2006-03-01

The design, commissioning and running of the reactor trip parameter monitoring system used in Qinshan Phase III are introduced. The applying technology of Distributed Control System realized trip parameter monitoring and realized the function of trip parameters quick data acquisitioning, transferring, saving, alarm, query. The applying of trip parameters monitoring system improved the abilities of plant status monitoring and event analyzing, and increased the security and economy of nuclear power plant. (authors)

Fuel Supply Shutdown Facility Interim Operational Safety Requirements

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

Use of an INGRES database to implement the beam parameter management at GANIL

International Nuclear Information System (INIS)

Gillette, P.; Lecorche, E.; Lermine, P.; Maugeais, C.; Leboucher, Ch.; Moscatello, M.H.; Pain, P.

1995-01-01

Since the beginning of the operation driven by the new Ganil control system in February 1993, the relational database management system (RDBMS) Ingres has been more and more widely used. The most significant application relying on the RDBMS is the new beam parameter management which has been entirely redesigned. It has been operational since the end of the machine shutdown in July this year. After a short recall of the use of Ingres inside the control system, the organization of the parameter management is presented. Then the database implementation is shown, including the way how the physical aspects of the Ganil tuning have been integrated in such an environment. (author)

78 FR 49553 - Three Mile Island, Unit 2; Post Shutdown Decommissioning Activities Report

Science.gov (United States)

2013-08-14

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-320; NRC-2013-0183] Three Mile Island, Unit 2; Post Shutdown Decommissioning Activities Report AGENCY: Nuclear Regulatory Commission. ACTION: Notice of receipt... Shutdown Decommissioning Activity Report (PSDAR) for Three Mile Island, Unit 2 (TMI-2). The PSDAR provides...

Startup, Shutdown, & Malfunction (SSM) Emissions at Industrial Facilities

Science.gov (United States)

EPA issued a final action to ensure states have plans in place that are fully consistent with the Clean Air Act and recent court decisions concerning startup, shutdown and malfunction (SSM) operations.

The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

2005-01-01

The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

Elementary calculation of the shutdown delay of a pile

International Nuclear Information System (INIS)

Yvon, J.

1949-04-01

This study analyzes theoretically the progress of the shutdown of a nuclear pile (reactor) when a cadmium rod is introduced instantaneously. For simplification reasons, the environment of the pile is considered as homogenous and only thermal neutrons are considered (delayed neutrons are neglected). Calculation is made first for a plane configuration (plane vessel, plane multiplier without reflector, and plane multiplier with reflector), and then for a cylindrical configuration (multiplier without reflector, multiplier with infinitely thick reflector, finite cylindrical piles without reflector and with reflector). The self-sustain conditions are calculated for each case and the multiplication length and the shutdown delay are deduced. (J.S.)

40 CFR 60.2685 - What happens during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 6 2010-07-01 2010-07-01 false What happens during periods of startup... happens during periods of startup, shutdown, and malfunction? (a) The emission limitations and operating limits apply at all times except during CISWI unit startups, shutdowns, or malfunctions. (b) Each...

40 CFR 60.3025 - What happens during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 6 2010-07-01 2010-07-01 false What happens during periods of startup... during periods of startup, shutdown, and malfunction? The emission limitations and operating limits apply at all times except during OSWI unit startups, shutdowns, or malfunctions. Model Rule—Performance...

Dependence of the time-constant of a fuel rod on different design and operational parameters

International Nuclear Information System (INIS)

Elenkov, D.; Lassmann, K.; Schubert, A.; Laar, J. van de

2001-01-01

The temperature response during a reactor shutdown has been measured for many years in the OECD-Halden Project. It has been shown that the complicated shutdown processes can be characterized by a time constant τ which depends on different fuel design and operational parameters, such as fuel geometry, gap size, fill gas pressure and composition, burnup and linear heat rate. In the paper the concept of a time constant is analyzed and the dependence of the time constant on various parameters is investigated analytically. Measured time constants for different designs and conditions are compared with those derived from calculations of the TRANSURANUS code. Employing standard models results in a systematic underprediction of the time constant, i.e. the heat transfer during shutdown is overestimated. (author)

Reload safety evaluation of boron dilution accident related to shutdown margin proportional to boron concentration

International Nuclear Information System (INIS)

Zee, Sung Kyun; Lee, Ki Bog; Song, Jae Woong

1993-06-01

This report investigates the efficient safety evaluation method and analysis procedure on Boron Dilution Accident(BDA) under the proportional shutdown margin to boron concentration. Also investigated are problems caused by applying this shutdown margin limit. Through this investigation, the safety of Kori-3 Cycle-8, Yonggwang-2 Cycle-7, Kori-4 Cycle-8 and Yonggwang-1 Cycle-8 with respect to BDA is verified. In order to satisfy the shutdown margin requirement in the Technical Specifications, it is shown that the High Flux Alarm at Shutdown Setting for Kori-4 Cycle-8 and Yonggwang-1 Cycle-8 at Mode 5 should be set at 2 or the Technical Specification should be revised. (Author)

40 CFR 60.1710 - What happens to the emission limits during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... during periods of startup, shutdown, and malfunction? 60.1710 Section 60.1710 Protection of Environment... during periods of startup, shutdown, and malfunction? (a) The emission limits of this subpart apply at all times except during periods of municipal waste combustion unit startup, shutdown, or malfunction...

40 CFR 60.1205 - What happens to the operating requirements during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... requirements during periods of startup, shutdown, and malfunction? 60.1205 Section 60.1205 Protection of... requirements during periods of startup, shutdown, and malfunction? (a) The operating requirements of this subpart apply at all times except during periods of municipal waste combustion unit startup, shutdown, or...

Shutdown and low-power operation at commercial nuclear power plants in the United States

International Nuclear Information System (INIS)

1993-09-01

The report contains the results of the NRC Staff's evaluation of shutdown and low-power operations at US commercial nuclear power plants. The report describes studies conducted by the staff in the following areas: Operating experience related to shutdown and low-power operations, probabilistic risk assessment of shutdown and low-power conditions and utility programs for planning and conducting activities during periods the plant is shut down. The report also documents evaluations of a number of technical issues regarding shutdown and low-power operations performed by the staff, including the principal findings and conclusions. Potential new regulatory requirements are discussed, as well as potential changes in NRC programs. A draft report was issued for comment in February 1992. This report is the final version and includes the responses to the comments along with the staff regulatory analysis of potential new requirements

40 CFR 62.14645 - What happens during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 8 2010-07-01 2010-07-01 false What happens during periods of startup... Limits § 62.14645 What happens during periods of startup, shutdown, and malfunction? (a) The emission limitations and operating limits apply at all times except during periods of CISWI unit startup, shutdown, or...

40 CFR 62.15165 - What happens to the emission limits during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... during periods of startup, shutdown, and malfunction? 62.15165 Section 62.15165 Protection of Environment... emission limits during periods of startup, shutdown, and malfunction? (a) The emission limits of this subpart apply at all times except during periods of municipal waste combustion unit startup, shutdown, or...

Extending reactor time-to-poison and reducing poison shutdown time by pre-shutdown power alterations

Energy Technology Data Exchange (ETDEWEB)

Kerr, Edward

1963-10-15

Manipulation of reactor power prior to shutdown and increasing the time- to-poison a sufficient amount to enable the required maintenance work to be completed and the reactor immediately restarted are discussed. The method employed in the NRU Reactor to gain the maximum timeto-poison with the least production loss is outlined. The method is based on intuition and is described by means of an analog of the iodine--xenon equations rather than the equations themselves. (C.E.S.)

Trends vs. reactor size of passive reactivity shutdown and control performance

International Nuclear Information System (INIS)

Wade, D.C.; Fujita, E.K.

1988-01-01

The focus of the US advanced reactor program since the cancellation of CRBR has been on inherent safety and cost reduction. The notion is to so design the reactor that in the event of an off normal condition, it brings itself to a safe shutdown condition and removes decay heat by reliance on ''inherent processes'' i.e., without reliance on devices requiring switching and outside sources of power. Such a reactor design would offer the potential to eliminate costly ''Engineered Safety Features,'' to lower capital costs, and to assuage public unease concerning reactor safety. For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the passive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. 8 refs., 12 figs., 1 tab

Development and validation of a model for high pressure liquid poison injection for CANDU-6 shutdown system no.2

International Nuclear Information System (INIS)

Rhee, B.-W.; Jeong, C.J.; Choi, J.H.; Yoo, S.-Y.

2002-01-01

In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the calandria tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, an AEA Technology CFD code, to simulate the formation and growth of the poison jet curtain inside the moderator tank. For validation, the current model is validated against a poison injection experiment performed at BARC, India and another poison jet experiment for Generic CANDU-6 performed at AECL, Canada. In conclusion this set of models is considered to predict the experimental results in a physically reasonable and consistent manner. (author)

An investigation of scramming the outer shutdown rods of the ANS with no reversal of flow in the manifold inlet lines

International Nuclear Information System (INIS)

Morsk, K.

1992-10-01

This report provides calculations and calculation checks on the outer shutdown system, consisting of eight shutdown rods located on the outside of the core. The function of the system is to scram the reactor, or to break the chain reaction of the fission process. The shutdown rods are clad with a neutron-absorbing material (i.e., hafnium) to achieve scram. During normal operation, the outer shutdown rods (Fig. 1) are in a nonscram, withdrawn position. This means that they are not close enough to the core to absorb a significant number of the neutrons that cause the fission process. In the case of a malfunction or an emergency, the outer control rods are moved to a position near the core. The outer shutdown system is operated with the use of springs and hydraulics. During normal operation, a constant flow of heavy water is circulated through the reflector vessel. A part of this flow provides a pressure high enough to keep the rods in their withdrawn or upper position, a nonscram status. If any signs of abnormal operation occur, the valves in the hydraulic system cut off the flow, and the springs push the rods into the scram position, stopping the chain reaction. Once the flow is restarted, the rods can be withdrawn to the nonscram position. Calculations of the mass of the outer control rod, the scram spring data, and the hydraulic pressure to hold the rods in the withdrawn position have been checked. In the case of a malfunction of the flow/pressure relief valves, a calculation was needed to show that the scram time would not exceed the time allowed. The scram time has been determined based on different values of the rod insertion length and the outside radius of the annulus was calculated. The effective force pushing the rod into the scram position, the rate of acceleration, and the actual scram time was then determined

Evaluation of reactivity shutdown margin for nuclear fuel reload optimization

International Nuclear Information System (INIS)

Wong, Hing-Ip; Maldonado, G.I.

1995-01-01

The FORMOSA-P code is a nuclear fuel management optimization package that combines simulated annealing (SA) and nodal generalized perturbation theory (GPT). Recent studies at Electricite de France (EdF-Clamart) have produced good results for power-peaking minimizations under multiple limiting control rod configurations. However, since the reactivity shutdown margin is not explicitly treated as an objective or constraint function, then any optimal loading patterns (LPs) are not guaranteed to yield an adequate shutdown margin (SDM). This study describes the implementation of the SDM calculation within a FORMOSA-P optimization. Maintaining all additional computational requirements to a minimum was a key consideration

Use of an INGRES database to implement the beam parameter management at GANIL

Energy Technology Data Exchange (ETDEWEB)

Gillette, P.; Lecorche, E.; Lermine, P.; Maugeais, C.; Leboucher, Ch.; Moscatello, M.H.; Pain, P.

1995-12-31

Since the beginning of the operation driven by the new Ganil control system in February 1993, the relational database management system (RDBMS) Ingres has been more and more widely used. The most significant application relying on the RDBMS is the new beam parameter management which has been entirely redesigned. It has been operational since the end of the machine shutdown in July this year. After a short recall of the use of Ingres inside the control system, the organization of the parameter management is presented. Then the database implementation is shown, including the way how the physical aspects of the Ganil tuning have been integrated in such an environment. (author). 2 refs.

Use of an INGRES database to implement the beam parameter management at GANIL

Energy Technology Data Exchange (ETDEWEB)

Gillette, P; Lecorche, E; Lermine, P; Maugeais, C; Leboucher, Ch; Moscatello, M H; Pain, P

1996-12-31

Since the beginning of the operation driven by the new Ganil control system in February 1993, the relational database management system (RDBMS) Ingres has been more and more widely used. The most significant application relying on the RDBMS is the new beam parameter management which has been entirely redesigned. It has been operational since the end of the machine shutdown in July this year. After a short recall of the use of Ingres inside the control system, the organization of the parameter management is presented. Then the database implementation is shown, including the way how the physical aspects of the Ganil tuning have been integrated in such an environment. (author). 2 refs.

Component failures that lead to manual shutdowns

International Nuclear Information System (INIS)

1979-01-01

The data for this report are taken from a population of thirty-five LWRs, al of which differ appreciably in size, design, and age. Appendix A provides a graphical display of the number of manual shutdowns per operating year as a function of plant age, with the frequency adjusted to reflect plant availability

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of inherent shutdown is emphasized in the approach to the design of innovative, small pool-type liquid-metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower events in evolving metal and oxide innovative designs

Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

International Nuclear Information System (INIS)

Mueller, C.J.

1986-01-01

The concept of ''inherent shutdown'' is emphasized in the approach to the design of innovative, small pool-type liquid metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram (ATWS) for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower (TOP) events in evolving metal and oxide innovative designs

Transient fission-product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

1997-12-01

Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

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

Loss of shutdown cooling during degassing in Doel 1

International Nuclear Information System (INIS)

1996-01-01

The presentation describes loss of shutdown cooling event during degassing in Doel 1 reactor, including description of Doel 1 features,status of plant prior to incident, event sequence and incident causes

Evaluation of reactivity shutdown margin for nuclear fuel reload optimization

International Nuclear Information System (INIS)

Engrand, P.; Wong, H. I.; Maldonado, G.I.

1996-01-01

The FORMOSA-P code is a nuclear fuel management optimization package which combines simulated annealing (SA) and nodal generalized perturbation theory (GPT). Recent studies at Electricite de France have produced good results for power peaking minimizations under multiple limiting control rod configurations. However, since the reactivity shutdown margin is not explicitly treated as an objective or constraint function, then any optimal loading patterns (LPs) are not guaranteed to yield an adequate shutdown margin (SDM). This study describes the implementation of the SDM calculation within a FORMOSA-P optimization. Maintaining all additional computational requirements to a minimum was a key consideration. (authors). 4 refs., 2 figs

Impact of shutdown risk on risk-based assessment of technical specifications

International Nuclear Information System (INIS)

Deriot, S.

1992-10-01

This paper describes the current work performed by the Research and Development Division of EDF concerning risk-based assessment of Operating Technical Specifications (OTS). The current risk-based assessment of OTS at EDF is presented. Then, the level 1 Probabilistic Safety Assessment of unit 3 of the Paluel nuclear power station (called PSA 1300) is described. It is fully computerized and takes into account the risk in shutdown states. A case study is presented. It shows that the fact of considering shutdown risk suggests that the current OTS should be modified

Type and timing of childhood maltreatment and severity of shutdown dissociation in patients with schizophrenia spectrum disorder.

Directory of Open Access Journals (Sweden)

Inga Schalinski

Full Text Available Dissociation, particularly the shutting down of sensory, motor and speech systems, has been proposed to emerge in susceptible individuals as a defensive response to traumatic stress. In contrast, other individuals show signs of hyperarousal to acute threat. A key question is whether exposure to particular types of stressful events during specific stages of development can program an individual to have a strong dissociative response to subsequent stressors. Vulnerability to ongoing shutdown dissociation was assessed in 75 inpatients (46 M/29 F, M = 31 ± 10 years old with schizophrenia spectrum disorder and related to number of traumatic events experienced or witnessed during childhood or adulthood. The Maltreatment and Abuse Chronology of Exposure (MACE scale was used to collect retrospective recall of exposure to ten types of maltreatment during each year of childhood. Severity of shutdown dissociation was related to number of childhood but not adult traumatic events. Random forest regression with conditional trees indicated that type and timing of childhood maltreatment could predictably account for 31% of the variance (p < 0.003 in shutdown dissociation, with peak vulnerability occurring at 13-14 years of age and with exposure to emotional neglect followed by various forms of emotional abuse. These findings suggest that there may be windows of vulnerability to the development of shutdown dissociation. Results support the hypothesis that experienced events are more important than witnessed events, but challenge the hypothesis that "life-threatening" events are a critical determinant.

Reserves for shutdown/dismantling and disposal in nuclear technology. Theses and recommendations on reform options

International Nuclear Information System (INIS)

Meyer, Bettina

2012-01-01

The study on reserves for shutdown, dismantling and disposal of nuclear facilities covers the following topics: cost for shutdown, dismantling and disposal and amount and transparency of nuclear reserves, solution by y stock regulated by public law for long-term liabilities, and improvement of the protection in the event of insolvency for the remaining EVU reserves for short- and intermediate-term liabilities. The appendix includes estimations and empirical values for the cost of shutdown and dismantling, estimation of disposal costs, and a summary of Swiss studies on dismantling and disposal and transfer to Germany.

Fuel supply shutdown facility interim operational safety requirements

International Nuclear Information System (INIS)

Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

1995-01-01

These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

300 Area fuel supply shutdown facility hazards assessment

International Nuclear Information System (INIS)

Campbell, L.R.

1998-01-01

This document establishes the technical basis in support of Emergency Planning activities for the 300 Area Fuel Supply Shutdown Facilities on the Hanford Site. Through this document, the technical basis for the development of facility specific Emergency Action Levels and Emergency Planning Zone, is demonstrated

Safeguards systems parameters

International Nuclear Information System (INIS)

Avenhaus, R.; Heil, J.

1979-01-01

In this paper analyses are made of the values of those parameters that characterize the present safeguards system that is applied to a national fuel cycle; those values have to be fixed quantitatively so that all actions of the safeguards authority are specified precisely. The analysis starts by introducing three categories of quantities: The design parameters (number of MBAs, inventory frequency, variance of MUF, verification effort and false-alarm probability) describe those quantities whose values have to be specified before the safeguards system can be implemented. The performance criteria (probability of detection, expected detection time, goal quantity) measure the effectiveness of a safeguards system; and the standards (threshold amount and critical time) characterize the magnitude of the proliferation problem. The means by which the values of the individual design parameters can be determined with the help of the performance criteria; which qualitative arguments can narrow down the arbitrariness of the choice of values of the remaining parameters; and which parameter values have to be fixed more or less arbitrarily, are investigated. As a result of these considerations, which include the optimal allocation of a given inspection effort, the problem of analysing the structure of the safeguards system is reduced to an evaluation of the interplay of only a few parameters, essentially the quality of the measurement system (variance of MUF), verification effort, false-alarm probability, goal quantity and probability of detection

Plant operational states analysis in low power and shutdown PSA

International Nuclear Information System (INIS)

He Jiandong; Qiu Yongping; Zhang Qinfang; An Hongzhen; Li Maolin

2013-01-01

The purpose of Plant Operational States (POS) analysis is to disperse the continuous and dynamic process of low power and shutdown operation, which is the basis of developing event tree models for accident sequence analysis. According to the design of a 300 MW Nuclear Power Plant Project, operating experience and procedures of the reference plant, a detailed POS analysis is carried out based on relative criteria. Then, several kinds of POS are obtained, and the duration of each POS is calculated according to the operation records of the reference plant. The POS analysis is an important element in low power and shutdown PSA. The methodology and contents provide reference for POS analysis. (authors)

Management of individual and collective dosimetry at Fessenheim nuclear plant. Evaluation after refueling shutdown

International Nuclear Information System (INIS)

Lamarre, D.; Waller, A.

1980-01-01

The principle of dosimetry management chosen by Fessenheim nuclear power station was originally consisted of two phases: - an automatic acquisition of individual doses realized by stylodosimeter readers; - a deferred data processing by computer. The whole system has not been used during the shutdown for the first refuelling of unit number one in view of encountered difficulties with perfecting of automatic readers prototype, this last phase has been replaced by a manual acquisition of doses. The dosimetry data processing has two main objects: - supervision of individual dosimetry for people who work in the nuclear power station; - knowledge of doses assigned for each working and equipment. Moreover, a first dosimetric result of the shutdown for refuelling of unit number one, enables to notice the workings which doses are the most important and written in percentage of total doses: regulatory controls: about 19%; - steam generators working: 16%; - working decontamination and making health physics screen (lock chamber) 10% [fr

Reactor shutdown device

Energy Technology Data Exchange (ETDEWEB)

Harada, Kiyoshi; Aono, Hidehiro [Hitachi Ltd., Tokyo (Japan); Fujita, Kaoru; Ishikawa, Tsuyoshi

1996-02-20

The present invention concerns a reactor shutdown device of a LMFBR type reactor, and provides a magnetic circuit having a sharp changing property of holding force relative to temperature change. Namely, a magnetic bridge is attached to a portion of the magnetic circuit. Then, required conditions are satisfied. Alternatively, even if the temperature dependent change of magnetic saturation of a temperature sensing alloy itself is somewhat moderated, the holding force from an erroneous dropping preventive temperature to a separating temperature can be abruptly reduced while keeping the holding force at a temperature lower than the erroneous dropping preventive temperature. Provision of the magnetic bridge increases the temperature dependent change of the holding force of the entire magnetic circuit. As a result, margin for the design of the temperature sensing alloy is extended. Actual design is enabled, and the range for selecting the temperature sensing alloy can be enlarged. (I.S.).

Reactor shutdown device

International Nuclear Information System (INIS)

Harada, Kiyoshi; Aono, Hidehiro; Fujita, Kaoru; Ishikawa, Tsuyoshi.

1996-01-01

The present invention concerns a reactor shutdown device of a LMFBR type reactor, and provides a magnetic circuit having a sharp changing property of holding force relative to temperature change. Namely, a magnetic bridge is attached to a portion of the magnetic circuit. Then, required conditions are satisfied. Alternatively, even if the temperature dependent change of magnetic saturation of a temperature sensing alloy itself is somewhat moderated, the holding force from an erroneous dropping preventive temperature to a separating temperature can be abruptly reduced while keeping the holding force at a temperature lower than the erroneous dropping preventive temperature. Provision of the magnetic bridge increases the temperature dependent change of the holding force of the entire magnetic circuit. As a result, margin for the design of the temperature sensing alloy is extended. Actual design is enabled, and the range for selecting the temperature sensing alloy can be enlarged. (I.S.)

Analysis of activation and shutdown contact dose rate for EAST neutral beam port

Science.gov (United States)

Chen, Yuqing; Wang, Ji; Zhong, Guoqiang; Li, Jun; Wang, Jinfang; Xie, Yahong; Wu, Bin; Hu, Chundong

2017-12-01

For the safe operation and maintenance of neutral beam injector (NBI), specific activity and shutdown contact dose rate of the sample material SS316 are estimated around the experimental advanced superconducting tokamak (EAST) neutral beam port. Firstly, the neutron emission intensity is calculated by TRANSP code while the neutral beam is co-injected to EAST. Secondly, the neutron activation and shutdown contact dose rates for the neutral beam sample materials SS316 are derived by the Monte Carlo code MCNP and the inventory code FISPACT-2007. The simulations indicate that the primary radioactive nuclides of SS316 are 58Co and 54Mn. The peak contact dose rate is 8.52 × 10-6 Sv/h after EAST shutdown one second. That is under the International Thermonuclear Experimental Reactor (ITER) design values 1 × 10-5 Sv/h.

Effects of shutdown chemistry on steam generator radiation levels at Point Beach Unit 2. Interim report

International Nuclear Information System (INIS)

Kormuth, J.W.

1982-05-01

A refueling shutdown chemistry test was conducted at a PWR, Point Beach Unit 2. The objective was to yield reactor coolant chemistry data during the cooldown/shutdown process which might establish a relationship between shutdown chemistry and its effects on steam generator radiation fields. Of particular concern were the effects of the presence of hydrogen in the coolant as contrasted to an oxygenated coolant. Analysis of reactor coolant samples showed a rapid soluble release (spike) in Co-58, Co-60, and nickel caused by oxygenation of the coolant. The measurement of radioisotope specific activities indicates that the material undergoing dissolution during the shutdown originated from different sources which had varying histories of activation. The test program developed no data which would support theories that oxygenation of the coolant while the steam generators are full of water contributes to increased steam generator radiation levels

An attempt for economic estimate of the shutdown of uranium production

International Nuclear Information System (INIS)

Jonchev, L.

1997-01-01

Uranium ore has been obtained since the end of 30s till 1992. No measures for protection of the environment and restricting the risk for the population during the production have been taken. Among the three possible models of shutting down the most inexpedient from economic point of view has been applied . It meant that the beginning of closing down took place far behind ceasing the production itself and the expenses for restoration were as big as fourteen times more in comparison to the two ones. The investments for prospecting and preparing new resources were lost. The whole process was made extremely inefficiently and unprofessionally. Because of the sudden closing down of production activities there was no enough time for gathering, processing and analyzing of necessary data, even the radioecological and hydro-ecological evaluations were doubtfully reliable. The shutdown of uranium production as worldwide practice takes place considering ALARA (As Low As Reasonably Achievable) principle. The aim is to achieve maximum possible results by minimum investments taking into account the radioecological risk, socially accounted for and psychologically conditioned expenses. There is no statement of the radioecological risk in the preliminary evaluations of the uranium mines in Bulgaria. The investment funds for the period 1992-1996 were about 2.1 bill. leva, (equally allocated for each year) which was about 46.5 mil. US$. Because of inflation process the investments crucially decreased during the last years when most capital-intensive activities had to be carried out - the engineering shutdown and land-reclamations procedures. The biggest share of investments (about 30 mil. US$) was for environmental status maintenance, 2.5 times less (about 13 mil. US$) - for technical shutdown and only 2.1 mil. US$ - for land reclamation. The investments for the shutdown process referred to the whole production obtained were only 2.5 US$/kg U 3 O 8 while the most effective model

Modelling the fluid structure interaction produced by a waterhammer during shutdown of high-pressure pumps

International Nuclear Information System (INIS)

Erath, W.; Nowotny, B.; Maetz, J.

1999-01-01

Measurements of an experiment in a pipe system with pump shutdown and valve closing have been performed in the nuclear power plant KRB II (Gundremmingen, Germany). Comparative calculations of fluid and structure including interaction show an excellent agreement with the measured results. Theory and implementation of the fluid structure interaction (FSI) and the results of the comparison are described. The following measurements have been compared with calculations: (1) experiments in Delft, Netherlands to analyse the FSI; and (2) experiment with pump shutdown and valve closing in the nuclear power plant KRB II has been performed. It turns out, that the consideration of the FSI is necessary for an exact calculation of 'soft' piping systems. It has significant application in current waterhammer problems. For example, water column closure, vapour collapse, check valve slamming continues to create waterhammers in the energy industry. An important consequence of the FSI is mostly a significant increase of the effective structural damping. This mitigates - so far in all KED's calculations the FSI has taken into account - an amplification of pipe movements due to pressure waves in resonance with structural eigenvalues. To investigate the integrity of pipe systems pipe stresses are calculated. Taking FSI into account they are reduced by 10-40% in the actual case. (orig.)

Inventory of radioactive corrosion products on the primary surfaces and release during shutdown in Ringhals 2

International Nuclear Information System (INIS)

Aronsson, O.

1994-01-01

In Ringhals 2 a retrospective study using gamma scans of system surfaces, fuel crud sampling and reactor coolant analyses during operation and shutdown has been done. The data have been used to prepare a balance of activity inventory. The inventory has been fairly stable from 1986 to 1993, expressed as a gamma source term. The steam generator replacement in 1989 removed some 40-50% of the Co-60 inventory in the reactor system. After the steam generator replacement, the gamma source term has got an increasing contribution from Co-58, absolutely as well as relatively. The reason for this is probably the switch from high pH operation to modified pH operation. Corrosion from fresh alloy 690 surfaces in the new steam generators is probably another contributing factor. The inventory and production rate of Co-60 is decreasing over the years. It has also been found that clean-up of the reactor coolant during start-up, operation, and shutdown as well as the fuel pool during refuelling removes about the same amounts of Co-60. (author). 11 figs., 15 refs

Technical Specification action statements requiring shutdown. A risk perspective with application to the RHR/SSW systems of a BWR

Energy Technology Data Exchange (ETDEWEB)

Mankamo, T. [Avaplan Oy, Espoo (Finland); Kim, I.S.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States)

1993-11-01

When safety systems fail during power operation, the limiting conditions for operation (LCOs) and associated action statements of technical specifications typically require that the plant be shut down within the limits of allowed outage time (AOT). However, when a system needed to remove decay heat, such as the residual heat removal (RHR) system, is inoperable or degraded, shutting down the plant may not necessarily be preferable, from a risk perspective, to continuing power operation over a usual repair time, giving priority to the repairs. The risk impact of the basic operational alternatives, i.e., continued operation or shutdown, was evaluated for failures in the RHR and standby service water (SSW) systems of a boiling-water reactor (BWR) nuclear power plant. A complete or partial failure of the SSW system fails or degrades not only the RHR system but other front-line safety systems supported by the SSW system. This report presents the methodology to evaluate the risk impact of LCOs and associated AOT; the results of risk evaluation from its application to the RHR and SSW systems of a BWR; the findings from the risk-sensitivity analyses to identify alternative operational policies; and the major insights and recommendations to improve the technical specifications action statements.

Shutdown dose rate contribution from diagnostics in ITER upper port 18

Energy Technology Data Exchange (ETDEWEB)

Cheon, M.S., E-mail: munseong@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Pak, S.; An, Y.H.; Seon, C.R.; Lee, H.G. [National Fusion Research Institute, Daejeon (Korea, Republic of); Bertalot, L.; Krasilnikov, V. [ITER Organization, St Paul-lez-Durance (France); Zvonkov, A. [Agency ITER-RF, Moscow (Russian Federation)

2016-11-01

Highlights: • The Shutdown Dose Rate in the interspace of ITER upper port 18 was evaluated. • VUV spectrometer is the dominant contributor to the average SDR. • The existence and size of the blanket cooling pipes impacts significantly on SDR. - Abstract: D-T operation of ITER plasma will produce high-energy fusion neutrons those can activate materials around the place where human-access is necessary. The interspace of the diagnostic port is one of the area where human-access is necessary for the maintenance of diagnostic systems installed at the port, so it is important to evaluate a dose rate of the interspace area in order to comply with ALARA principle. The shutdown dose rate (SDR) in the interspace of ITER upper port 18 was evaluated by the Direct 1-Step (D1S) method using MCNP5 code. This port contains three diagnostics: Vacuum Ultra-Violet (VUV) Spectrometer, Neutron Activation System (NAS), and Upper Vertical Neutron Camera (UVNC). The contribution of each diagnostic in the port was evaluated by running separate upper port MCNP models those contain individual diagnostic only, and the total dose rate contribution was evaluated with the model which was fully integrated with all the diagnostics. The effect of the opening around the upper port plug and of the other ports was also investigated. The purpose of this assessment is to provide the shielding design basis for the preliminary design of the diagnostic integration in the port. The method and result of the calculation will be presented in this paper.

Review of occupational radiation exposures in all biennial shutdown maintenance of Kaiga generating station

International Nuclear Information System (INIS)

Murukan, E.K.; Vinod Kumar, T.; Austine, N.X.; Soumia Menon, M.; Girish Kumar, K.; Rao, M.M.L.N.; Venkataramana, K.

2008-01-01

Full text: Kaiga generating station 1 and 2 consists of twin units of 220 M We pressurized heavy water reactors located in Karnataka, India. Major maintenance activities of one of the twin units are taken up once in two years (biennial shutdown) to execute system maintenance, system up gradation, surveillance and in-service inspection (ISI) jobs. BSDs are mandatory activities to comply with regulatory requirement to ensure the safety and reliability of plant system equipment. More than 65% of the station collective dose is contributed by biennial shutdown (BSD) jobs. It is observed that the man rem consumed during normal operation of the plant is less than 35% of the total man rem consumed. Since BSD jobs contributes significantly to station collective dose, an effective implementation of radiation protection programme specific to BSD is the key to control the occupational exposure. Various improvements in the field of radiation protection practices and process systems are adopted to achieve lowest collective dose at par with international standards. The key areas identified for application of various strategies to achieve ALARA were Man rem budgeting, Radiological condition monitoring, Radiation protection practices, Identification of critical jobs and Work groups, Work planning and execution, and Radioactive waste management. Review of collective doses of all the BSD jobs performed in the station since year 2004 and various measures incorporated to achieve ALARA exposures to plant personnel are briefly discussed in this paper. (author)

Method of disposing of shut-down nuclear power plants

International Nuclear Information System (INIS)

Gaiser, H.

1984-01-01

A shut-down atomic power plant or a section thereof, particularly the nuclear reactor, is disposed of by sinking it to below ground level by constructing a caisson with cutting edges from the foundations of said plant or section or by excavating a pit therebelow

The Chernobyl plant shutdown

International Nuclear Information System (INIS)

2000-12-01

The Chernobylsk-1 reactor, operational in september 1977 has been stopped in november 1996; the Chernobylsk-2 reactor started in november 1978 is out of order since 1991 following a fire. The Chernobylsk-3 reactor began in 1981. During the last three years it occurs several maintenance operations that stop it. In june 2000, the Ukrainian authorities decided to stop it definitively on the 15. of december (2000). This file handles the subject. it is divided in four chapters: the first one gives the general context of the plant shutdown, the second chapter studies the supporting projects to stop definitively the nuclear plant, the third chapter treats the question of the sarcophagus, and the fourth and final chapter studies the consequences of the accident and the contaminated territories. (N.C.)

On the startup and shutdown of a tandem mirror reactor

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.T.; Fisher, J.L.; Madden, P.A.

1979-01-01

The startup and shutdown of a fusion reactor must be performed in such a way that the plasma remains MHD stable. In a tandem mirror the stability depends on a sufficiently high pressure ratio between the plugs and the central cell, of the order of 100. Control of the neutral beam input to the plugs by means of active feedback has been investigated to achieve an acceptable pressure ratio throughout the entire startup/shutdown transient. An algorithm to control the beam input power has been developed. The control law was subsequently tested in a tandem mirror simulation code. This paper describes the basic models incorporated in the simulation, as well as the derivation of the control algorithm. The simulation results are presented and the practicality of implementing the algorithm is discussed. 4 refs

Modeling startup and shutdown transient of the microlinear piezo drive via ANSYS

Science.gov (United States)

Azin, A. V.; Bogdanov, E. P.; Rikkonen, S. V.; Ponomarev, S. V.; Khramtsov, A. M.

2017-02-01

The article describes the construction-design of the micro linear piezo drive intended for a peripheral cord tensioner in the reflecting surface shape regulator system for large-sized transformable spacecraft antenna reflectors. The research target -the development method of modeling startup and shutdown transient of the micro linear piezo drive. This method is based on application software package ANSYS. The method embraces a detailed description of the calculation stages to determine the operating characteristics of the designed piezo drive. Based on the numerical solutions, the time characteristics of the designed piezo drive are determined.

Trends vs. reactor size of passive reactivity shutdown and control performance

International Nuclear Information System (INIS)

Wade, D.C.; Fujita, E.K.

1987-01-01

For LMR concepts, the goal of passive reactivity shutdown has been approached in the US by designing the reactors for favorable relationships among the power, power/flow, and inlet temperature coefficients of reactivity, for high internal conversion ratio (yielding small burnup control swing), and for a primary pump coastdown time appropriately matched to the delayed neutron hold back of power decay upon negative reactivity input. The use of sodium bonded metallic fuel pins has facilitated the achievement of the massive shutdown design goals as a consequence of their high thermal conductivity and high effective heavy metal density. Alternately, core designs based on derated oxide pins may be able to achieve the passive shutdown features at the cost of larger core volume and increased initial fissile inventory. For LMR concepts, the passive decay heat removal goal of inherent safety has been approached in US designs by use of pool layouts, larger surface to volume ratio of the reactor vessel with natural draft air cooling of the vessel surface, elevations and redans which promote natural circulation through the core, and thermal mass of the pool contents sufficient to absorb that initial transient decay heat which exceeds the natural draft air cooling capacity. This paper describes current US ''inherently safe'' reactor design

An SBLOCA Test for Shutdown Cooling Line Break Using the SMART-ITL Facility

Energy Technology Data Exchange (ETDEWEB)

Bae, Hwang; Kim, Dong Eok; Ryu, Sung Uk; Shin, Yong Cheol; Ko, Yung Joo; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The objectives of SMART-ITL are to investigate and understand the integral performance of the reactor systems and components, and the thermalhydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. Its height was preserved and its area and volume were scaled down to 1/49 compared with the SMART prototype plant. The SMART-ITL consists of a primary system including a reactor pressure vessel with a pressurizer, four steam generators and four main coolant pumps, a secondary system, a safety system, and an auxiliary system. The SMART was installed at KAERI and several transient tests were recently finished. In this paper, the test results for a steady-state operation and a transient of the small break loss of coolant accident (SBLOCA) are discussed. An SBLOCA test simulating the shutdown cooling line break was performed using SMART-ITL properly. All parameters were in good agreement with the target values during the steady-state operation period. The pressures and temperatures show reasonable behaviors during the SBLOCA test. SMART (System-integrated Modular Advanced ReacTor) which was designed by KAERI is an integral type reactor. The standard design approval for the SMART design was issued on July 4th of 2012 by a Korean regulatory body, the Nuclear Safety and Security Commission (NSSC). The main components including a pressurizer, steam generators, and reactor coolant pumps are installed in a reactor pressure vessel, and there are no large-size pipes. The safety systems could be simplified as an LBLOCA (Large-Break Loss of Coolant Accident) scenario is inherently excluded. An integral-effect test loop for SMART (SMART-ITL, or FESTA) was designed to simulate the integral thermal-hydraulic behavior of SMART. The SMART-ITL has been designed using a volume scaling methodology.

Shielding optimisation of the ITER ICH&CD antenna for shutdown dose rate

International Nuclear Information System (INIS)

Turner, Andrew; Leichtle, Dieter; Lamalle, Philippe; Levesy, Bruno; Meunier, Lionel; Polunovskiy, Eduard; Sartori, Roberta; Shannon, Mark

2015-01-01

Highlights: • Neutronics analysis on the ITER ICH&CD system conducted to reduce shutdown dose rate. • Several designs for shielding the port plug gaps were modelled. • Shielding significantly reduced interspace dose rate but still exceed project requirements. • Design optimisation of the ICH port is continuing. • Significant contributions from other ports require an integrated modelling approach. - Abstract: The Ion Cyclotron Heating and Current Drive (ICH&CD) system will reside in ITER equatorial port plugs 13 and 15. Shutdown dose rates (SDDR) within the port interspace are required to be less than 100 μSv/h at 10 6 s cooling. A significant contribution to the SDDR results from neutrons streaming down gaps around the port frame, and the mitigation of this streaming is the main subject of these analyses. An updated MCNP model of the antenna was created and integrated into an ITER reference model. Shielding plates were defined in the port gaps, and scoping studies conducted to assess their effectiveness in several configurations, based on which a front dog-leg arrangement was selected for high resolution 3-D activation analysis using MCR2S. It was concluded that the selected configuration reduced the SDDR from ∼500 μSv/h to 220 μSv/h but were still in excess of dose rate requirements. Approximately 30% of this was due to cross-talk from neighbouring ports. In addition, increased dose rates were observed in the port interspace along the lines of sight of the removable vacuum transmission lines. Design optimisation is continuing, however an integrated approach is needed with regard to ITER port plug design and the shielding of surrounding systems.

Shielding optimisation of the ITER ICH&CD antenna for shutdown dose rate

Energy Technology Data Exchange (ETDEWEB)

Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Leichtle, Dieter [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Lamalle, Philippe; Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St., Paul-lez-Durance (France); Meunier, Lionel [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Polunovskiy, Eduard [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St., Paul-lez-Durance (France); Sartori, Roberta [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Shannon, Mark [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2015-10-15

Highlights: • Neutronics analysis on the ITER ICH&CD system conducted to reduce shutdown dose rate. • Several designs for shielding the port plug gaps were modelled. • Shielding significantly reduced interspace dose rate but still exceed project requirements. • Design optimisation of the ICH port is continuing. • Significant contributions from other ports require an integrated modelling approach. - Abstract: The Ion Cyclotron Heating and Current Drive (ICH&CD) system will reside in ITER equatorial port plugs 13 and 15. Shutdown dose rates (SDDR) within the port interspace are required to be less than 100 μSv/h at 10{sup 6} s cooling. A significant contribution to the SDDR results from neutrons streaming down gaps around the port frame, and the mitigation of this streaming is the main subject of these analyses. An updated MCNP model of the antenna was created and integrated into an ITER reference model. Shielding plates were defined in the port gaps, and scoping studies conducted to assess their effectiveness in several configurations, based on which a front dog-leg arrangement was selected for high resolution 3-D activation analysis using MCR2S. It was concluded that the selected configuration reduced the SDDR from ∼500 μSv/h to 220 μSv/h but were still in excess of dose rate requirements. Approximately 30% of this was due to cross-talk from neighbouring ports. In addition, increased dose rates were observed in the port interspace along the lines of sight of the removable vacuum transmission lines. Design optimisation is continuing, however an integrated approach is needed with regard to ITER port plug design and the shielding of surrounding systems.

Safety and regulation aspects of nuclear facilities shutdown

International Nuclear Information System (INIS)

Clement, B.

1977-01-01

Technical dispositions that safety authorities will accept after shutdown of a nuclear installation and reglementation to use are examined. The different solutions from surveillance and maintenance, after removal of fissile materials and radioactive fluids, to dismantling are discussed especially for reactors. In each case the best solution has to be studied to ensure protection of public health and environment [fr

Questions and answers about the reactor shutdown at the Barsebaeck plant

International Nuclear Information System (INIS)

1992-01-01

At a scram at the Barsebaeck 2 reactor on July 28 1992, a safety valve open unintentionally, and steam was released from the reactor vessel into the containment. The emergency spray system started sprinkling the vessel (the core spray system was also active for a short while). After one hour, the sprinkling was interupted, and at about the same time it was found that the steam jet had tore off insulation material (from the containment walls) which started to clog the sieves for the emergency sprinkling water, disturbing the pumping. The clogging appeared much more rapidly than expected (1 h in stead of 10 h). Five Swedish reactors for similar design have been shutdown pending a reconstruction of the emergency spray feed system. This pamphlet is directed to the general public, explaining the problems and commenting on nuclear safety issues

Role of tumor necrosis factor in flavone acetic acid-induced tumor vasculature shutdown

International Nuclear Information System (INIS)

Mahadevan, V.; Malik, S.T.; Meager, A.; Fiers, W.; Lewis, G.P.; Hart, I.R.

1990-01-01

Flavone acetic acid (FAA), a novel investigational antitumor agent, has been shown to cause early vascular shutdown in several experimental murine tumors, and this phenomenon is believed to be crucial to FAA's antitumor effects. However, the basis of this FAA-induced tumor vascular shutdown is unknown. In this study a radioactive tracer-clearance technique has been used as an objective indication of tumor blood flow to show that i.p. administered FAA induces a progressive and sustained reduction in blood flow in a colon 26 tumor growing s.c. in syngeneic mice. As early as 1 h after administration, there was a significant increase in the t1/2 clearance value for intratumorally injected 133Xe, reaching a peak at 3 h (117.3 +/- 36.4 versus 7.8 +/- 0.85 min for controls). Significant inhibition of blood flow was still apparent 48 h after a single injection of drug. This FAA-induced vascular shutdown was virtually abolished in tumor-bearing mice pretreated with an antiserum against tumor necrosis factor, while no such effect was observed in controls pretreated with nonimmune serum (t1/2 of 10.8 +/- 1.2 versus 65.6 +/- 8.0 min for controls). Furthermore, in vitro FAA was seen to induce tumor necrosis factor secretion from murine peritoneal cells and splenocytes. These studies suggest that FAA-induced tumor vascular shutdown in the colon 26 tumor is mediated by tumor necrosis factor

235U Holdup Measurement Program in support of facility shutdown

International Nuclear Information System (INIS)

Thomason, R.S.; Griffin, J.C.; Lien, O.G.; McElroy, R.D.

1991-01-01

In 1989, the Department of Energy directed shutdown of an enriched uranium processing facility at Savannah River Site. As part of the shutdown requirements, deinventory and cleanout of process equipment and nondestructive measurement of the remaining 235 U holdup were required. The holdup measurements had safeguards, accountability, and nuclear criticality safety significance; therefore, a technically defensible and well-documented holdup measurement program was needed. Appropriate standards were fabricated, measurement techniques were selected, and an aggressive schedule was followed. Early in the program, offsite experts reviewed the measurement program, and their recommendations were adopted. Contact and far-field methods were used for most measurements, but some process equipment required special attention. All holdup measurements were documented, and each report was subjected to internal peer review. Some measured values were checked against values obtained by other methods; agreement was generally good

Evaluation of nuclear parameters measurements realized during the Angra-1 reactor start-up tests and the theoretical previsions of these parameters

International Nuclear Information System (INIS)

Fernandes, V.B.; Ponzoni Filho, P.; Perrotta, J.A.; Silva Ipojuca, T. da

1982-01-01

A summary of measuring techniques and the results of the follow nuclear parameters, are presented: a) reactivity of various control bank; b) capacity of reactor shutdown; c) critical concentrations of soluble boron for various core configurations; d) reactivity isothermal coefficients; e) power mapping; f) power reactivity coefficients. (E.G.) [pt

Understanding the hydrologic impacts of wastewater treatment plant discharge to shallow groundwater: Before and after plant shutdown

Science.gov (United States)

Hubbard, Laura E.; Keefe, Steffanie H.; Kolpin, Dana W.; Barber, Larry B.; Duris, Joseph W.; Hutchinson, Kasey J.; Bradley, Paul M.

2016-01-01

Effluent-impacted surface water has the potential to transport not only water, but wastewater-derived contaminants to shallow groundwater systems. To better understand the effects of effluent discharge on in-stream and near-stream hydrologic conditions in wastewater-impacted systems, water-level changes were monitored in hyporheic-zone and shallow-groundwater piezometers in a reach of Fourmile Creek adjacent to and downstream of the Ankeny (Iowa, USA) wastewater treatment plant (WWTP). Water-level changes were monitored from approximately 1.5 months before to 0.5 months after WWTP closure. Diurnal patterns in WWTP discharge were closely mirrored in stream and shallow-groundwater levels immediately upstream and up to 3 km downstream of the outfall, indicating that such discharge was the primary control on water levels before shutdown. The hydrologic response to WWTP shutdown was immediately observed throughout the study reach, verifying the far-reaching hydraulic connectivity and associated contaminant transport risk. The movement of WWTP effluent into alluvial aquifers has implications for potential WWTP-derived contamination of shallow groundwater far removed from the WWTP outfall.

Causes of extended shutdown state of 'RA' research reactor in Vinca Institute

International Nuclear Information System (INIS)

Pesic, M.; Kolundzija, V.; Ljubenov, V.; Cupac, S.

2001-01-01

This paper describes the causes and reasons for extended shutdown state of RA research reactor in the 'Vinca' Institute of Nuclear Sciences. Technical and legal matters that led to decision to stop RA reactor operation in 1984 and further problems related to maintenance and preparation for continuation of operation are given. Influence of nuclear policy of Yugoslav government and the 'Vinca' Institute at prolongation of the reactor shutdown state, as consequence of changing of nuclear programme in the country and the world are discussed and underlined. An overview of the legislation in the field of nuclear safety and regulatory control of radiation sources and radioactive materials in Yugoslavia is presented. (author)

Does debt ceiling and government shutdown help in forecasting the us equity risk premium?

Directory of Open Access Journals (Sweden)

Aye Goodness C.

2016-01-01

Full Text Available This article evaluates the predictability of the equity risk premium in the United States by comparing the individual and complementary predictive power of macroeconomic variables and technical indicators using a comprehensive set of 16 economic and 14 technical predictors over a monthly out-ofsample period of 1995:01 to 2012:12 and an in-sample period of 1986:01- 1994:12. In order to do so we consider, in addition to the set of variables used in Christopher J. Neely et al. (2013 and using a more recent dataset, the forecasting ability of two other important variables namely government shutdown and debt ceiling. Our results show that one of the newly added variables namely government shutdown provides statistically significant out-of-sample predictive power over the equity risk premium relative to the historical average. Most of the variables, including government shutdown, also show significant economic gains for a risk averse investor especially during recessions.

Summary of Session 5 and 6 'Long Shutdown 1'

Energy Technology Data Exchange (ETDEWEB)

Bordry, F; Foraz, K [European Organization for Nuclear Research, Geneva (Switzerland)

2012-07-01

This paper summarizes the sessions devoted to Long Shutdown 1 (LS1) in the LHC, injectors and experiments. The time frame and start date were discussed, with the main activities from powering tests prior to warm-up up to physics were presented. The session finished with a discussion on the maximum reasonable energy. (author)

Analysis of Gamma Dose Rate Caused by Corrosion Products inside the Containment Building of Yonngwang Nuclear Power Plant Unit 3 During Shutdown Period

Energy Technology Data Exchange (ETDEWEB)

Ha, Wi Ho; Kim, Jae Cheon; Kim, Soon Young; Kim, Jong Kyung [Hanyang Univ., Seoul (Korea, Republic of)

2005-07-01

Occupational radiation exposure(ORE) of nuclear power plant(NPP) workers mainly occurs during the shutdown period. Major radioactive sources are the corrosion products released from the reactor coolant system(RCS). The corrosion products consist of circulating crud and deposited crud. Major radioactive corrosion products, {sup 58}Co and {sup 60}Co, are known to contribute approximately more than 70% of the total ORE. In this study, the corrosion products regarding cobalt were evaluated during the shutdown period, and gamma dose rates caused by them were calculated at the main working area inside the containment building of the Yonggwang NPP Unit 3.

The Shutdown Dissociation Scale (Shut-D)

Science.gov (United States)

Schalinski, Inga; Schauer, Maggie; Elbert, Thomas

2015-01-01

The evolutionary model of the defense cascade by Schauer and Elbert (2010) provides a theoretical frame for a short interview to assess problems underlying and leading to the dissociative subtype of posttraumatic stress disorder. Based on known characteristics of the defense stages “fright,” “flag,” and “faint,” we designed a structured interview to assess the vulnerability for the respective types of dissociation. Most of the scales that assess dissociative phenomena are designed as self-report questionnaires. Their items are usually selected based on more heuristic considerations rather than a theoretical model and thus include anything from minor dissociative experiences to major pathological dissociation. The shutdown dissociation scale (Shut-D) was applied in several studies in patients with a history of multiple traumatic events and different disorders that have been shown previously to be prone to symptoms of dissociation. The goal of the present investigation was to obtain psychometric characteristics of the Shut-D (including factor structure, internal consistency, retest reliability, predictive, convergent and criterion-related concurrent validity). A total population of 225 patients and 68 healthy controls were accessed. Shut-D appears to have sufficient internal reliability, excellent retest reliability, high convergent validity, and satisfactory predictive validity, while the summed score of the scale reliably separates patients with exposure to trauma (in different diagnostic groups) from healthy controls. The Shut-D is a brief structured interview for assessing the vulnerability to dissociate as a consequence of exposure to traumatic stressors. The scale demonstrates high-quality psychometric properties and may be useful for researchers and clinicians in assessing shutdown dissociation as well as in predicting the risk of dissociative responding. PMID:25976478

The Shutdown Dissociation Scale (Shut-D

Directory of Open Access Journals (Sweden)

Inga Schalinski

2015-05-01

Full Text Available The evolutionary model of the defense cascade by Schauer and Elbert (2010 provides a theoretical frame for a short interview to assess problems underlying and leading to the dissociative subtype of posttraumatic stress disorder. Based on known characteristics of the defense stages “fright,” “flag,” and “faint,” we designed a structured interview to assess the vulnerability for the respective types of dissociation. Most of the scales that assess dissociative phenomena are designed as self-report questionnaires. Their items are usually selected based on more heuristic considerations rather than a theoretical model and thus include anything from minor dissociative experiences to major pathological dissociation. The shutdown dissociation scale (Shut-D was applied in several studies in patients with a history of multiple traumatic events and different disorders that have been shown previously to be prone to symptoms of dissociation. The goal of the present investigation was to obtain psychometric characteristics of the Shut-D (including factor structure, internal consistency, retest reliability, predictive, convergent and criterion-related concurrent validity.A total population of 225 patients and 68 healthy controls were accessed. Shut-D appears to have sufficient internal reliability, excellent retest reliability, high convergent validity, and satisfactory predictive validity, while the summed score of the scale reliably separates patients with exposure to trauma (in different diagnostic groups from healthy controls.The Shut-D is a brief structured interview for assessing the vulnerability to dissociate as a consequence of exposure to traumatic stressors. The scale demonstrates high-quality psychometric properties and may be useful for researchers and clinicians in assessing shutdown dissociation as well as in predicting the risk of dissociative responding.

Development of advanced automatic operation system for nuclear ship. 1. Perfect automatic normal operation

International Nuclear Information System (INIS)

Nakazawa, Toshio; Yabuuti, Noriaki; Takahashi, Hiroki; Shimazaki, Junya

1999-02-01

Development of operation support system such as automatic operating system and anomaly diagnosis systems of nuclear reactor is very important in practical nuclear ship because of a limited number of operators and severe conditions in which receiving support from others in a case of accident is very difficult. The goal of development of the operation support systems is to realize the perfect automatic control system in a series of normal operation from the reactor start-up to the shutdown. The automatic control system for the normal operation has been developed based on operating experiences of the first Japanese nuclear ship 'Mutsu'. Automation technique was verified by 'Mutsu' plant data at manual operation. Fully automatic control of start-up and shutdown operations was achieved by setting the desired value of operation and the limiting value of parameter fluctuation, and by making the operation program of the principal equipment such as the main coolant pump and the heaters. This report presents the automatic operation system developed for the start-up and the shutdown of reactor and the verification of the system using the Nuclear Ship Engineering Simulator System. (author)

Reactor shutdown device

International Nuclear Information System (INIS)

Ito, Masahiko

1990-01-01

The object of the present invention is to reliably shutdown an LMFBR type reactor upon accident of the reactor. That is, curie point magnetic member is made annular so that it can be moved between the outer circumference of an electromagnet and the position above the electromagnet. This enables to enlarge the curie point magnetic member since it is no more necessary to be inserted it in a guide tube. Accordingly, attracting force upon normal operation is increased to remarkably improve the reliability against erronerous scram, etc. Further, since a required gap is formed between the curie point magnetic member and the electromagnet and the heat of coolants is efficiently transmitted to the curie point magnetic member, rapid scram is possible. Further, a position support mechanism is disposed to a part of a control element or at the inner side of the guiding tube for urging and actuating the armature to make it protrude above the top of the guiding tube. With such a constitution, since the armature can be adsorbed without inserting the curie point magnetic member and the electromagnet guide tube, the same effect as in the case of inserting them can be obtained. (I.S.)

Replacement energy, capacity, and reliability costs for permanent nuclear reactor shutdowns

International Nuclear Information System (INIS)

VanKuiken, J.C., Buehring, W.A.; Hamilton, S.; Kavicky, J.A.; Cavallo, J.D.; Veselka, T.D.; Willing, D.L.

1993-10-01

Average replacement power costs are estimated for potential permanent shutdowns of nuclear electricity-generating units. Replacement power costs are considered to include replacement energy, capacity, and reliability cost components. These estimates were developed to assist the US Nuclear Regulatory Commission in evaluating regulatory issues that potentially affect changes in serious reactor accident frequencies. Cost estimates were derived from long-term production-cost and capacity expansion simulations of pooled utility-system operations. Factors that affect replacement power cost, such as load growth, replacement sources of generation, and capital costs for replacement capacity, were treated in the analysis. Costs are presented for a representative reactor and for selected subcategories of reactors, based on estimates for 112 individual reactors

Guidance of reactor operators and TSC personnel with the severe accident management guidance under shutdown and low power conditions

International Nuclear Information System (INIS)

Van Haesendonck, M.F.; Prior, R.P.

2000-01-01

The Westinghouse Owners Group Severe Accident Management Guidance (WOG SAMG) was developed between 1991 and 1994. The primary goals for severe accident management that form the basis of the WOG SAMG are to terminate any radioactive releases to the environment; to prevent failure of any containment fission product boundary and to return the plant to a controlled stable condition. The WOG SAMG is primarily a TSC tool for mitigation of low probability core damage events. The philosophy is that control room operators should remain focused on the prevention of core damage, whereas the TSC personnel should concentrate on the mitigation of the severe accident. The symptom based package is built up as a structured process for choosing appropriate actions based on actual plant conditions. No detailed knowledge of severe accident phenomena is required. The scope of the WOG SAMG is limited to severe accidents resulting from initiating events occurring during full power operation. However, a number of studies such as the EdF EPS 1300 Probabilistic Safety Assessment (PSA), the shutdown Probabilistic Risk Assessment (PRA) for Surry, the BERA shutdown PRA for Beznau, the EPRI/ Westinghouse ORAM methodology etc. have shown that the frequency of core damage (a severe accident) during shutdown and low power operation can be of the same order of magnitude as for full power operation. The at-power SAMG is viewed as the resolution of the severe accident issue. Similarly, it is expected that as shutdown PRAs mature, the final resolution of the severe accident issue will lie in SAMG for low power and shutdown operation. Therefore in resolution of this issue, Westinghouse has developed the Shutdown Severe Accident Management Guidance (SSAMG) which gives guidance for both control room and TSC personnel to mitigate a severe accident under shutdown or low power conditions. In the last few years, many LWR plants have been implementing SAMG. In the US, all plants have developed SAMG, and many

Evolution of the ATLAS Distributed Computing during the LHC long shutdown

CERN Document Server

Campana, S; The ATLAS collaboration

2013-01-01

The ATLAS Distributed Computing project (ADC) was established in 2007 to develop and operate a framework, following the ATLAS computing model, to enable data storage, processing and bookkeeping on top of the WLCG distributed infrastructure. ADC development has always been driven by operations and this contributed to its success. The system has fulfilled the demanding requirements of ATLAS, daily consolidating worldwide up to 1PB of data and running more than 1.5 million payloads distributed globally, supporting almost one thousand concurrent distributed analysis users. Comprehensive automation and monitoring minimized the operational manpower required. The flexibility of the system to adjust to operational needs has been important to the success of the ATLAS physics program. The LHC shutdown in 2013-2015 affords an opportunity to improve the system in light of operational experience and scale it to cope with the demanding requirements of 2015 and beyond, most notably a much higher trigger rate and event pileu...

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Massaro, Lawrence M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-09-30

A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power reactor sites was conducted. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

Francis-99: Transient CFD simulation of load changes and turbine shutdown in a model sized high-head Francis turbine

International Nuclear Information System (INIS)

Mössinger, Peter; Jester-Zürker, Roland; Jung, Alexander

2017-01-01

With increasing requirements for hydropower plant operation due to intermittent renewable energy sources like wind and solar, numerical simulations of transient operations in hydraulic turbo machines become more important. As a continuation of the work performed for the first workshop which covered three steady operating conditions, in the present paper load changes and a shutdown procedure are investigated. The findings of previous studies are used to create a 360° model and compare measurements with simulation results for the operating points part load, high load and best efficiency. A mesh motion procedure is introduced, allowing to represent moving guide vanes for load changes from best efficiency to part load and high load. Additionally an automated re-mesh procedure is added for turbine shutdown to ensure reliable mesh quality during guide vane closing. All three transient operations are compared to PIV velocity measurements in the draft tube and pressure signals in the vaneless space. Simulation results of axial velocity distributions for all three steady operation points, during both load changes and for the shutdown correlated well with the measurement. An offset at vaneless space pressure is found to be a result of guide vane corrections for the simulation to ensure similar velocity fields. Short-time Fourier transformation indicating increasing amplitudes and frequencies at speed-no load conditions. Further studies will discuss the already measured start-up procedure and investigate the necessity to consider the hydraulic system dynamics upstream of the turbine by means of a 1D3D coupling between the 3D flow field and a 1D system model. (paper)

Francis-99: Transient CFD simulation of load changes and turbine shutdown in a model sized high-head Francis turbine

Science.gov (United States)

Mössinger, Peter; Jester-Zürker, Roland; Jung, Alexander

2017-01-01

With increasing requirements for hydropower plant operation due to intermittent renewable energy sources like wind and solar, numerical simulations of transient operations in hydraulic turbo machines become more important. As a continuation of the work performed for the first workshop which covered three steady operating conditions, in the present paper load changes and a shutdown procedure are investigated. The findings of previous studies are used to create a 360° model and compare measurements with simulation results for the operating points part load, high load and best efficiency. A mesh motion procedure is introduced, allowing to represent moving guide vanes for load changes from best efficiency to part load and high load. Additionally an automated re-mesh procedure is added for turbine shutdown to ensure reliable mesh quality during guide vane closing. All three transient operations are compared to PIV velocity measurements in the draft tube and pressure signals in the vaneless space. Simulation results of axial velocity distributions for all three steady operation points, during both load changes and for the shutdown correlated well with the measurement. An offset at vaneless space pressure is found to be a result of guide vane corrections for the simulation to ensure similar velocity fields. Short-time Fourier transformation indicating increasing amplitudes and frequencies at speed-no load conditions. Further studies will discuss the already measured start-up procedure and investigate the necessity to consider the hydraulic system dynamics upstream of the turbine by means of a 1D3D coupling between the 3D flow field and a 1D system model.

25 CFR 226.28 - Shutdown, abandonment, and plugging of wells.

Science.gov (United States)

2010-04-01

... OSAGE RESERVATION LANDS FOR OIL AND GAS MINING Cessation of Operations § 226.28 Shutdown, abandonment... production of oil and/or gas has been demonstrated to the satisfaction of the Superintendent. Lessee shall... the means by which the well bore is to be protected, and the contemplated eventual disposition of the...

Control rod shutdown system

International Nuclear Information System (INIS)

Miyamoto, Yoshiyuki; Higashigawa, Yuichi.

1996-01-01

The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)

Safety shutdowns and failures of the RA reactor equipment; Sigurnosna zaustavljanja i kvarovi opreme na reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Mitrovic, S [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

1966-07-01

This report is an attempt of statistical analysis of the failures occurred during RA reactor operation. A list of failures occurred on the RA equipment during 1965 is included. Failures were related to the following systems: dosimetry system (22%), safety and control system (7%), heavy water system (2%), technical water (4%), helium system (2%), measuring instruments (30%), transport, ventilation, power supply systems (32%). A review of safety shutdowns from 1962 to 1966 is included as well, as a comparison with three similar reactors. Although the number of events used for statistical analysis was not adequate, it has been concluded that RA reactor operation was stable and reliable.

Medical surveillance of nuclear power plant workers during reactor shutdown using whole-body counting and excretion analysis

International Nuclear Information System (INIS)

Le Roux-Desmis, C.

1987-01-01

After a review of radioactivity basis and radiation protection principles, the various aspects of medical surveillance of nuclear power plant workers during reactor shutdown, are presented. Internal contamination incidents that happened during 1986-1987 shutdown of Paluel reactor are exposed. Internal contamination levels are evaluated using whole-body counting and radionuclide determination in feces and urine and compared with dose limits [fr

Adjustment of the Kompleks Titan-2 monitoring computerized system at the Rovno NPP third unit

International Nuclear Information System (INIS)

Zigaev, B.P.

1987-01-01

Information signal origin and processing processes in the monitoring computerized system 'Komplex Titan-2' at the Rovno NPP third unit are considered. The system exercises control over the following production equipment parameters: the state of keys of control of lock fittings, mechanisms, regulators, reserve mechanism automatic shut-down circuit; the state of lock fittings, mechanisms, regulators, reserve mechanism automated switching on; parameter deviation from permissible values; interlock operation; protection system state; the state of autonomous units and devices

Order concerning a nuclear reactor shutdown

International Nuclear Information System (INIS)

Anon.

1991-01-01

Judgment of the State Administrative Court of Baden Wuerttemberg in head notes including: The authority of the Minister-President to give general guidelines includes the right to issue single directives; in matters of prime political significance he can take measures to realize such aims. - It is no extraneous consideration for the supervisory board under atomic energy law to point out in an order concerning a nuclear reactor shutdown that the disallowed operation of a nuclear plant conflicts with the obligation of the state to provide protection and constitutes a penal offence. Further a discourse on the assignment of discretionary powers under Paragraph 19 Section 3 Clause 2 No. 3 of the Atomic Energy Law. (HSCH) [de

A parameter tree approach to estimating system sensitivities to parameter sets

International Nuclear Information System (INIS)

Jarzemba, M.S.; Sagar, B.

2000-01-01

A post-processing technique for determining relative system sensitivity to groups of parameters and system components is presented. It is assumed that an appropriate parametric model is used to simulate system behavior using Monte Carlo techniques and that a set of realizations of system output(s) is available. The objective of our technique is to analyze the input vectors and the corresponding output vectors (that is, post-process the results) to estimate the relative sensitivity of the output to input parameters (taken singly and as a group) and thereby rank them. This technique is different from the design of experimental techniques in that a partitioning of the parameter space is not required before the simulation. A tree structure (which looks similar to an event tree) is developed to better explain the technique. Each limb of the tree represents a particular combination of parameters or a combination of system components. For convenience and to distinguish it from the event tree, we call it the parameter tree. To construct the parameter tree, the samples of input parameter values are treated as either a '+' or a '-' based on whether or not the sampled parameter value is greater than or less than a specified branching criterion (e.g., mean, median, percentile of the population). The corresponding system outputs are also segregated into similar bins. Partitioning the first parameter into a '+' or a '-' bin creates the first level of the tree containing two branches. At the next level, realizations associated with each first-level branch are further partitioned into two bins using the branching criteria on the second parameter and so on until the tree is fully populated. Relative sensitivities are then inferred from the number of samples associated with each branch of the tree. The parameter tree approach is illustrated by applying it to a number of preliminary simulations of the proposed high-level radioactive waste repository at Yucca Mountain, NV. Using a

Low Power Shutdown PSA for CANDU Type Plants

Energy Technology Data Exchange (ETDEWEB)

Bae, Yeon Kyoung; Kim, Myung Su [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

KHNP also have concentrated on full power PSA. Some recently constructed OPR1000 type plants and APR1400 type plants have performed the low power and shutdown (LPSD) PSA. The purpose of LPSD PSA is to identify the main contributors on the accident sequences of core damage and to find the measure of safety improvement. After the Fukushima accident, Korean regulatory agency required the shutdown severe accident management guidelines (SSAMG) development for safety enhancement. For the reliability of SSAMG, KHNP should develop the LPSD PSA. Especially, the LPSD PSA for CANDU type plant had developed for the first time in Korea. This paper illustrates how the LPSD PSA for CANDU type developed and the core damage frequency (CDF) is different with that of full power PSA. KHNP performed LPSD PSA to develop the SSAMG after the Fukushima accidents. The results show that risk at the specific operation mode during outage is higher than that of full power operation. Also, the results indicated that recovery failure of class 4 power at the POS 5A, 5B contribute dominantly to the total CDF from importances analysis. LPSD PSA results such as CDF with initiating events and POSs, risk results with plant damage state, and containment failure probability and frequency with POSs can be used by inputs for developing the SSAMG.

Runaway electron generation during plasma shutdown by killer pellet injection

International Nuclear Information System (INIS)

Gal, K; Feher, T; Smith, H; Fueloep, T; Helander, P

2008-01-01

Tokamak discharges are sometimes terminated by disruptions that may cause large mechanical and thermal loads on the vessel. To mitigate disruption-induced problems it has been proposed that 'killer' pellets could be injected into the plasma in order to safely terminate the discharge. Killer pellets enhance radiative energy loss and thereby lead to rapid cooling and shutdown of the discharge. But pellets may also cause runaway electron generation, as has been observed in experiments in several tokamaks. In this work, runaway dynamics in connection with deuterium or carbon pellet-induced fast plasma shutdown is considered. A pellet code, which calculates the material deposition and initial cooling caused by the pellet is coupled to a runaway code, which determines the subsequent temperature evolution and runaway generation. In this way, a tool has been created to test the suitability of different pellet injection scenarios for disruption mitigation. If runaway generation is avoided, the resulting current quench times are too long to safely avoid large forces on the vessel due to halo currents

Parameters in pure type systems

NARCIS (Netherlands)

Bloo, C.J.; Kamareddine, F.; Laan, T.D.L.; Nederpelt, R.P.; Rajsbaum, S.

2002-01-01

In this paper we study the addition of parameters to typed ¿-calculus with definitions. We show that the resulting systems have nice properties and illustrate that parameters allow for a better fine-tuning of the strength of type systems as well as staying closer to type systems used in practice in

Report of a consultants meeting on accidents during shutdown conditions for WWER nuclear power plants. Extrabudgetary programme on the safety of WWER NPPs

International Nuclear Information System (INIS)

1996-07-01

The main objectives of the meeting were to exchange information on the operational occurrences, studies performed and countermeasures taken for the accidents during shutdown for WWERs, and to define the necessity and directions of the further activities which may promote the improvement of WWER safety under shutdown conditions. The consultants have discussed some aspects concerning vulnerability of safety functions during shutdown conditions, several steps required to performed accident analysis and selected operational aspects for shutdown conditions. The discussion was supported by an evaluation of selected operational occurrences. The consultants have agreed that the discussion during the meeting in major parts is relevant to all the WWER designs (i.e. WWER-1000, WWER-440/213 and WWER-440/230). As for the plant conditions, the consultants have agreed to bound the discussion mainly by the cold shutdown and refuelling modes. Refs, figs, tabs

Reactor shut-down device

International Nuclear Information System (INIS)

Otsuka, Fumio; Horikawa, Yuji.

1990-01-01

The present invention concerns an externally disposed reactor shut-down device for an FBR type reactor using liquid sodium as coolants. An introducing pipe having an outlet port disposed at an upper portion thereof is disposed at a lower end of an upper guide tube. An extension tube, an L-shaped measuring wire support and a measuring wire are disposed at the inside of the guide tube. With such a constitution, low temperature coolants flown out from the lower guide tube of a control rod and a great amount of high temperature coolants flown out from the lower guide tube of a fuel assembly are introduced smoothly to the introducing tube having the measuring wire support disposed therein. Accordingly, the high temperature coolants can be prevented from flowing out to the outside of the introducing tube and coolants after mixing can be flown and hit against a curie point electromagnet efficiently. This can make the response to abnormal temperature rise of coolants satisfactory and can provide reliable reactor scram. (I.N.)

Evaluation of induced activity, decay heat and dose rate distribution after shutdown in ITER

Energy Technology Data Exchange (ETDEWEB)

Maki, Koichi [Hitachi Ltd., Ibaraki (Japan). Hitachi Research Lab.; Satoh, Satoshi; Hayashi, Katsumi; Yamada, Koubun; Takatsu, Hideyuki; Iida, Hiromasa

1997-03-01

Induced activity, decay heat and dose rate distributions after shutdown were estimated for 1MWa/m{sup 2} operation in ITER. The activity in the inboard blanket one day after shutdown is 1.5x10{sup 11}Bq/cm{sup 3}, and the average decay heating rate 0.01w/cm{sup 3}. The dose rate outside the 120cm thick concrete biological shield is two order higher than the design criterion of 5{mu}Sv/h. This indicates that the biological shield thickness should be enhanced by 50cm in concrete, that is, total thickness 170cm for workers to enter the reactor room and to perform maintenance. (author)

Method for calculating the forces and deformations in the fast reactor fuel assembly accounting for the effects of reactor control system elements and shutdown

International Nuclear Information System (INIS)

Likhachev, Yu.I.; Vashlyaev, Yu.N.; Kravchenko, I.N.

1980-01-01

Methods for calculating deformations and interaction forces of heat-generating assemblies (HGA) of fast reactor core with account for the effect of control and protection system (CPS) elements at the reactor operation and change of interaction efforts between HGA at the reactor shutdown, are described. The results of testing the suggested methods on example of estimate of HGA behaviour of the BN-350 reactor are presented. For estimating the effect of CPS elements on HGA bending the sector model has been used. It is assumed that HGA deformation inside each sector is independent of HGA deformation of other sectors. A higher calculation accuracy is attained by means of laying out of sectors into regions of preferable influence of emergency protection elements and compensating packets. When determining deformation and interaction efforts between HGA caused by temperature change in the course of shutdown it is supposed that the HGA deformation is purely elastic. The methods described are realized in the form of ABRI-CPS and ABRI-HOL programs written in FORTRAN for the BESM-6 computer. The results of HGA calculations of the BN-350 reactor core show that CPS elements decrease contact efforts in the middle of the central packet, increase contact efforts in the peak of the central packet, increase contact efforts in the peaks of packets from the eight row to the periphery and increase contact efforts in the middles of packets from the 5th to 9th row [ru

Outcomes of an international initiative for harmonization of low power and shutdown probabilistic safety assessment

Directory of Open Access Journals (Sweden)

Manna Giustino

2010-01-01

Full Text Available Many probabilistic safety assessment studies completed to the date have demonstrated that the risk dealing with low power and shutdown operation of nuclear power plants is often comparable with the risk of at-power operation, and the main contributors to the low power and shutdown risk often deal with human factors. Since the beginning of the nuclear power generation, human performance has been a very important factor in all phases of the plant lifecycle: design, commissioning, operation, maintenance, surveillance, modification, decommissioning and dismantling. The importance of this aspect has been confirmed by recent operating experience. This paper provides the insights and conclusions of a workshop organized in 2007 by the IAEA and the Joint Research Centre of the European Commission, on Harmonization of low power and shutdown probabilistic safety assessment for WWER nuclear power plants. The major objective of the workshop was to provide a comparison of the approaches and the results of human reliability analyses and gain insights in the enhanced handling of human factors.

Telemetry System of Biological Parameters

Directory of Open Access Journals (Sweden)

Jan Spisak

2005-01-01

Full Text Available The mobile telemetry system of biological parameters serves for reading and wireless data transfer of measured values of selected biological parameters to an outlying computer. It concerns basically long time monitoring of vital function of car pilot.The goal of this projects is to propose mobile telemetry system for reading, wireless transfer and processing of biological parameters of car pilot during physical and psychical stress. It has to be made with respect to minimal consumption, weight and maximal device mobility. This system has to eliminate signal noise, which is created by biological artifacts and disturbances during the data transfer.

Estimation of shutdown heat generation rates in GHARR-1 due to ...

African Journals Online (AJOL)

Fission products decay power and residual fission power generated after shutdown of Ghana Research Reactor-1 (GHARR-1) by reactivity insertion accident were estimated by solution of the decay and residual heat equations. A Matlab program code was developed to simulate the heat generation rates by fission product ...

Study on the performance of the Particle Identification Detectors at LHCb after the LHC First Long Shutdown (LS1)

CERN Document Server

Fontana, Marianna

2016-01-01

During the First Long Shutdown (LS1), the LHCb experiment has introduced major modification in the data-processing procedure and modified part of the detector to deal with the increased energy and the increased heavy-hadron production cross-section. In this contribution we review the performance of the particle identification detectors at LHCb, Rich, Calorimeters, and Muon system, after the LS1

Shutdown dose rate analysis with CAD geometry, Cartesian/tetrahedral mesh, and advanced variance reduction

International Nuclear Information System (INIS)

Biondo, Elliott D.; Davis, Andrew; Wilson, Paul P.H.

2016-01-01

Highlights: • A CAD-based shutdown dose rate analysis workflow has been implemented. • Cartesian and superimposed tetrahedral mesh are fully supported. • Biased and unbiased photon source sampling options are available. • Hybrid Monte Carlo/deterministic techniques accelerate photon transport. • The workflow has been validated with the FNG-ITER benchmark problem. - Abstract: In fusion energy systems (FES) high-energy neutrons born from burning plasma activate system components to form radionuclides. The biological dose rate that results from photons emitted by these radionuclides after shutdown—the shutdown dose rate (SDR)—must be quantified for maintenance planning. This can be done using the Rigorous Two-Step (R2S) method, which involves separate neutron and photon transport calculations, coupled by a nuclear inventory analysis code. The geometric complexity and highly attenuating configuration of FES motivates the use of CAD geometry and advanced variance reduction for this analysis. An R2S workflow has been created with the new capability of performing SDR analysis directly from CAD geometry with Cartesian or tetrahedral meshes and with biased photon source sampling, enabling the use of the Consistent Adjoint Driven Importance Sampling (CADIS) variance reduction technique. This workflow has been validated with the Frascati Neutron Generator (FNG)-ITER SDR benchmark using both Cartesian and tetrahedral meshes and both unbiased and biased photon source sampling. All results are within 20.4% of experimental values, which constitutes satisfactory agreement. Photon transport using CADIS is demonstrated to yield speedups as high as 8.5·10"5 for problems using the FNG geometry.

Shutdown risk management applied at Philadelphia Electric Company

International Nuclear Information System (INIS)

Dagan, William J.; True, Douglas E.; Wilson, Thomas; Truax, William

2004-01-01

The development and implementation of an effective risk management program requires basic risk or safety knowledge and the conversion of such information into effective management tools. ERIN Engineering and Research, Inc., under contract to the Electric Power Research Institute, has developed an effective program. Outage Risk Assessment and Management (ORAM), to provide plant and management personnel with understandable results of shutdown risk studies. With this tool, the impact of plans and decision options can be readily determined and displayed for the decision maker. This paper describes these methods and their application to the Limerick Nuclear Station of Philadelphia Electric Company. It also sets forth a broader application of these methods to include support of management decisions at-power and following forced outages. The result is an integrated risk management framework which can allow management and technical personnel to utilize readily available and understandable risk insights to optimize each activity. This paper addresses the resolution of several key issues in detail: How was the ORAM risk management method employed to represent the existing plant shutdown procedures and policies? How did the ORAM risk management method enhance the decision-making ability of the outage management staff? How was the ORAM software efficiently integrated with the outage scheduling software? How is quantitative risk information generated and used for outage planning and control? The ORAM risk management philosophy utilizes a series of colors to depict various risk configurations. Each such configuration has associated with it clear guidance. By modifying the conditions existing in the plant it is possible to impact the type of risk being encountered as well as the guidance which is appropriate for that period. In addition, the duration of a particular configuration can be effectively managed to reduce the overall risk impact. These are achieved with minimal

Small leak shutdown, location, and behavior in LMFBR steam generators

International Nuclear Information System (INIS)

Sandusky, D.W.

1976-01-01

The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

Dynamic Analysis of a Floating Vertical Axis Wind Turbine Under Emergency Shutdown Using Hydrodynamic Brake

DEFF Research Database (Denmark)

Wang, K.; Hansen, Martin Otto Laver; Moan, T.

2014-01-01

Emergency shutdown is always a challenge for an operating vertical axis wind turbine. A 5-MW vertical axis wind turbine with a Darrieus rotor mounted on a semi-submersible support structure was examined in this study. Coupled non-linear aero-hydro-servo-elastic simulations of the floating vertical...... axis wind turbine were carried out for emergency shutdown cases over a range of environmental conditions based on correlated wind and wave data. When generator failure happens, a brake should be applied to stop the acceleration of the rotor to prevent the rotor from overspeeding and subsequent disaster...

Impact of Government Shutdown on Child Care and Early Education Programs

Science.gov (United States)

Center for Law and Social Policy, Inc. (CLASP), 2013

2013-01-01

Congress did not enact a continuing resolution bill by midnight September 30, 2013, thereby triggering a partial government shutdown effective October 1, 2013. October 1 began the federal fiscal year 2014. Most discretionary programs, those that are subject to the annual Congressional appropriations process, will not receive 2014 funding. Most,…

Hazard Classification for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

Final hazard classification for the 300 Area N Reactor fuel storage facility resulted in the assignment of Nuclear Facility Hazard Category 3 for the uranium metal fuel and feed material storage buildings (303-A, 303-B, 303-G, 3712, and 3716). Radiological for the residual uranium and thorium oxide storage building and an empty former fuel storage building that may be used for limited radioactive material storage in the future (303-K/3707-G, and 303-E), and Industrial for the remainder of the Fuel Supply Shutdown buildings (303-F/311 Tank Farm, 303-M, 313-S, 333, 334 and Tank Farm, 334-A, and MO-052)

Program of social protection for Chornobyl nuclear power plant staff and Slavutich town residents in the aftermath of the plant shutdown

International Nuclear Information System (INIS)

Komarov, V.A.

2001-01-01

In order to solve social issues related to ChNPP shutdown, the Ukrainian Government approved 'Program of Social Protection for Chornobyl Nuclear Power Plant Staff and Slavutich Town Residents in Aftermath of Plant Shutdown' on 29 November 2000. The Program Objective is to ensure social protection and support of well being of ChNPP staff and Slavutich town residents after the plant shutdown. Preserve and develop town infrastructure. Create compensatory jobs; efficiently manage human resources; provide social allowances and guarantees to the ChNPP staff that is being released, and Slavutich town residents

Importance theory for lumped-parameter systems

International Nuclear Information System (INIS)

Cady, K.B.; Kenton, M.A.; Ward, J.C.; Piepho, M.G.

1981-01-01

A general sensitivity theory has been developed for nonlinear lumped parameter system simulations. The point of departure is general perturbation theory for nonlinear systems. Importance theory as developed here allows the calculation of the sensitivity of a response function to any physical or design parameter; importance of any equation or term or physical effect in the system model on the response function; variance of the response function caused by the variances and covariances of all physical parameters; and approximate effect on the response function of missing physical phenomena or incorrect parameters

0-d modeling of fast radiative shutdown of Tokamak discharges following massive gas injection

International Nuclear Information System (INIS)

Hollmann, E.M.; Parks, P.B.; Scott, H.A.

2008-01-01

0-D modeling of fast radiative shutdowns of tokamak discharges following massive gas injection is presented. Realistic neutral deposition rates are used together with a 1-D diffusive model to estimate impurity deposition into the plasma. Non-coronal radiation rates including opacity are used, as are induced wall currents, wall impurity radiation, and neutral and neoclassical corrections to plasma resistivity. The 0-D modeling is found to reproduce the shutdown timescale and free electron density rise seen in DIII-D argon injection experiments well. Opacity, wall currents, and wall impurities can all have a significant (>10%) impact on simulated timescales. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Performance Parameters for Grid-Connected PV Systems

Energy Technology Data Exchange (ETDEWEB)

Marion, B.; Adelstein, J.; Boyle, K.; Hayden, H.; Hammond, B.; Fletcher, T.; Canada, B.; Narang, D.; Shugar, D.; Wenger, H.; Kimber, A.; Mitchell, L.; Rich, G.; Townsend, T.

2005-02-01

The use of appropriate performance parameters facilitates the comparison of grid-connected photovoltaic (PV) systems that may differ with respect to design, technology, or geographic location. Four performance parameters that define the overall system performance with respect to the energy production, solar resource, and overall effect of system losses are the following: final PV system yield, reference yield, performance ratio, and PVUSA rating. These performance parameters are discussed for their suitability in providing desired information for PV system design and performance evaluation and are demonstrated for a variety of technologies, designs, and geographic locations. Also discussed are methodologies for determining system a.c. power ratings in the design phase using multipliers developed from measured performance parameters.The use of appropriate performance parameters facilitates the comparison of grid-connected photovoltaic (PV) systems that may differ with respect to design, technology, or geographic location. Four performance parameters that define the overall system performance with respect to the energy production, solar resource, and overall effect of system losses are the following: final PV system yield, reference yield, performance ratio, and PVUSA rating. These performance parameters are discussed for their suitability in providing desired information for PV system design and performance evaluation and are demonstrated for a variety of technologies, designs, and geographic locations. Also discussed are methodologies for determining system a.c. power ratings in the design phase using multipliers developed from measured performance parameters.

On the functional failures concept and probabilistic safety margins: challenges in application for evaluation of effectiveness of shutdown systems - 15318

International Nuclear Information System (INIS)

Serghiuta, D.; Tholammakkil, J.

2015-01-01

The use of level-3 reliability approach and the concept of functional failure probability could provide the basis for defining a safety margin metric which would include a limit for the probability of functional failure, in line with the definition of a reliability-based design. It can also allow a quantification of level of confidence, by explicit modeling and quantification of uncertainties, and provide a better framework for representation of actual design and optimization of design margins within an integrated probabilistic-deterministic model. This paper reviews the attributes and challenges in application of functional failure concept in evaluation of risk-informed safety margins using as illustrative example the case of CANDU reactors shutdown systems effectiveness. A risk-informed formulation is first introduced for estimation of a reasonable limit for the functional failure probability using a Swiss cheese model. It is concluded that more research is needed in this area and a deterministic - probabilistic approach may be a reasonable intermediate step for evaluation of functional failure probability at the system level. The views expressed in this paper are those of the authors and do not necessarily reflect those of CNSC, or any part thereof. (authors)

Behavior of ruthenium in the case of shutdown of the cooling system of HLLW storage tanks

International Nuclear Information System (INIS)

Philippe, M.; Mercier, J.P.; Gue, J.P.

1990-01-01

The consequences of the failure of the cooling system of fission product storage tanks over a variable period were investigated as part of the safety analysis of the La Hague spent fuel reprocessing plant. Due to the considerable heat release, induced by the fission products, a prolonged shutdown of the tank cooling system could cause the progressive evaporation of the solutions to dryness, and culminate in the formation of volatile species of ruthenium and their release in the tank venting circuit. To determine the fraction of ruthenium likely to be transferred from the storage tanks in volatile or aerosol form during the failure, evaporation tests were conducted by evaporating samples of actual nitric acid solutions of fission products, obtained on the laboratory scale after the reprocessing of several kilograms of MOX fuels irradiated to 30,000 MWday.t -1 . A distillation apparatus was designed to operate with small volume solution samples, reproducing the heating conditions existing in the reprocessing plant within a storage tank for fission products. The main conclusions drawn from these experiments are as follows: - ruthenium is only volatilized in the final phase of evaporation, just before desiccation, - for a final temperature limited to 160 0 C, the total fraction of volatilized ruthenium reaches 12%, in the presence of H 2 0, HN0 3 , N0 x and 0 2 , the volatilized ruthenium recombines mainly in the form of ruthenium nitrosyl nitrates, or decomposes into ruthenium oxide on the walls of the apparatus. Assuming a heating power density of 10 W/liter of concentrate, and a perfectly adiabatic storage system, the minimum time required to reach dryness can be estimated at 90 h, allowing substantial time to take action to restore a cooling source

Behavior of ruthenium in the case of shutdown of the cooling system of HLLW storage tanks

International Nuclear Information System (INIS)

Philippe, M.; Mercier, J.P.; Gue, J.P.

1991-01-01

The consequences of the failure of the cooling system of fission product storage tanks over a variable period were investigated as part of the safety analysis of the La Hague spent fuel reprocessing plant. Due to the considerable heat release, induced by the fission products, a prolonged shutdown of the tank cooling system could cause the progressive evaporation of the solutions to dryness, and culminate in the formation of volatile species of ruthenium and their release in the tank venting circuit. To determine the fraction of ruthenium likely to be transferred from the storage tanks in volatile or aerosol form during the failure, evaporation tests were conducted by evaporating samples of actual nitric acid solutions of fission products, obtained on the laboratory scale after the reprocessing of several kilograms of MOX fuels irradiated to 30,000 MW day·t -1 . A distillation apparatus was designed to operate with small-volume solution samples, reproducing the heating conditions existing in the reprocessing plant within a storage tank for fission products. The main conclusions drawn from these experiments are as follows: ruthenium is only volatilized in the final phase of evaporation, just before desiccation; for a final temperature limited to 160 degree C, the total fraction of volatilized ruthenium reaches 12%; in the presence of H 2 O, HNO 3 , NO x and O 2 , the volatilized ruthenium recombines mainly in the form of ruthenium nitrosyl nitrates, or decomposes into ruthenium oxide (probably RuO 2 ) on the walls of the apparatus. Assuming a heating power density of 10 W/liter of concentrate, and a perfectly adiabatic storage system, the minimum time required to reach dryness can be estimated at 90 h, allowing substantial time to take action to restore a cooling source

On the Interplay between Order Parameter Dynamics and System Parameter Dynamics in Human Perceptual-Cognitive-Behavioral Systems.

Science.gov (United States)

Frank, T D

2015-04-01

Previous research has demonstrated that perceiving, thinking, and acting are human activities that correspond to self-organized patterns. The emergence of such patterns can be completely described in terms of the dynamics of the pattern amplitudes, which are referred to as order parameters. The patterns emerge at bifurcations points when certain system parameters internal and external to a human agent exceed critical values. At issue is how one might study the order parameter dynamics for sequences of consecutive, emergent perceptual, cognitive, or behavioral activities. In particular, these activities may in turn impact the system parameters that have led to the emergence of the activities in the first place. This interplay between order parameter dynamics and system parameter dynamics is discussed in general and formulated in mathematical terms. Previous work that has made use of this two-tiered framework of order parameter and system parameter dynamics are briefly addressed. As an application, a model for perception under functional fixedness is presented. Finally, it is argued that the phenomena that emerge in this framework and can be observed when human agents perceive, think, and act are just as likely to occur in pattern formation systems of the inanimate world. Consequently, these phenomena do not necessarily have a neurophysiological basis but should instead be understood from the perspective of the theory of self-organization.

Psychoacoustic parameters and its measuring system; Onshitsu hyoka wo hyokasuru tame no parameter to keisoku system

Energy Technology Data Exchange (ETDEWEB)

Ohashi, M.; Imaizumi, H.; Ono, T. [Ono Sokki Co. Ltd., Tokyo (Japan)

1998-05-01

Human auditory sensation has both extremely excellent performance and general versatility as sound analyzer. At present, it is impossible to make equipment with the same functions as human being, and describe an auditory sensation function as acoustic sensor even by any physical analysis techniques. However, extraction of auditory sensation parameters is becoming possible by using psychoacoustics and binaural signal processing. This paper mainly explains the calculation method of sound quality evaluation parameters derived from psychoacoustic results based on a sound quality evaluation system under development by the authors. This system is based on binaural measurement by dummy head, and calculates psychoacoustic parameters such as loudness, sharpness, roughness, fluctuation strength and tonality through frequency analysis of the measured stereo signals. The system also calculates 2-D parameters such as sensory pleasantness and unbiased annoyance based on the above parameters. 12 refs., 4 figs.

46 CFR 32.50-35 - Remote manual shutdown for internal combustion engine driven cargo pump on tank vessels-TB/ALL.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Remote manual shutdown for internal combustion engine... for Cargo Handling § 32.50-35 Remote manual shutdown for internal combustion engine driven cargo pump on tank vessels—TB/ALL. (a) Any tank vessel which is equipped with an internal combustion engine...

Knowledge-based time management in the planning of plant shut-downs; WISENT: Wissensbasiertes Zeitmanagement bei der Planung von Anlagenstillstaenden

Energy Technology Data Exchange (ETDEWEB)

Verweyen-Frank, H.

1996-12-31

At STEAG of Essen, an operator of hard-coal-fired power stations, the WISENT system distributes the necessary shut-down times in order to permit precautionary maintenance work to be carried out. This expert system, developed in cooperation with the company ALBIT, has a particular trait. It is not only based on ``if``-statements, as is typical of expert systems, but makes use also of fuzzy methods in order to optimize overall distribution. In other words, this system is the first to add to the strict yes-or-no logic a weighting component for each criterion of optimization. (orig.)

Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

International Nuclear Information System (INIS)

Rivas, J.C.; Dies, J.; Fajarnés, X.

2015-01-01

Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

Despite the Shutdown, Rescheduled NIH Research Festival Brings Science to the Forefront | Poster

Science.gov (United States)

By Andrea Frydl, Contributing Writer Although it was delayed by almost a month because of the federal shutdown, the NIH Research Festival still took place at the NIH Clinical Center in Bethesda, Md., and attendance was high.

Enhancing Efficiency of Safeguards at Facilities that are Shutdown or Closed-Down, including those being Decommissioned

Energy Technology Data Exchange (ETDEWEB)

Moran, B. [Brookhaven National Lab. (BNL), Upton, NY (United States); Stern, W. [Brookhaven National Lab. (BNL), Upton, NY (United States); Colley, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Marzo, M. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2016-12-15

International Atomic Energy Agency (IAEA) safeguards involves verification activities at a wide range of facilities in a variety of operational phases (e.g., under construction, start-up, operating, shutdown, closed-down, and decommissioned). Safeguards optimization for each different facility type and operational phase is essential for the effectiveness of safeguards implementation. The IAEA’s current guidance regarding safeguards for the different facility types in the various lifecycle phases is provided in its Design Information Examination (DIE) and Verification (DIV) procedure. 1 Greater efficiency in safeguarding facilities that are shut down or closed down, including those being decommissioned, could allow the IAEA to use a greater portion of its effort to conduct other verification activities. Consequently, the National Nuclear Security Administration’s Office of International Nuclear Safeguards sponsored this study to evaluate whether there is an opportunity to optimize safeguards approaches for facilities that are shutdown or closed-down. The purpose of this paper is to examine existing safeguards approaches for shutdown and closed-down facilities, including facilities being decommissioned, and to seek to identify whether they may be optimized.

Simulating fission product transients via the history-based local-parameter methodology

International Nuclear Information System (INIS)

Jenkins, D.A.; Rouben, B.; Salvatore, M.

1993-01-01

This paper describes the fission-product-calculation capacity of the history-based local-parameter methodology for evaluating lattice properties for use in core-tracking calculations in CANDU reactors. In addition to taking into account the individual past history of each bundles flux/power level, fuel temperature, and coolant density and temperature that the bundle has seen during its stay in the core, the latest refinement of the history-based method provides the capability of fission-product-drivers. It allows the bundle-specific concentrations of the three basic groups of saturating fission products to be calculated in steady state or following a power transient, including long shutdowns. The new capability is illustrated by simulating the startup period following a typical long-shutdown, starting from a snapshot in the Point Lepreau operating history. 9 refs., 7 tabs

Performance of Resistive Plate Chambers installed during the first long shutdown of the CMS experiment

CERN Document Server

Shopova, M.; Aleksandrov, A.; Hadjiiska, R.; Iaydjiev, P.; Sultanov, G.; Rodozov, M.; Stoykova, S.; Assran, Y.; Sayed, A.; Radi, A.; Aly, S.; Singh, G.; Abbrescia, M.; Iaselli, G.; Maggi, M.; Pugliese, G.; Verwilligen, P.; Van Doninck, W.; Colafranceschi, S.; Sharma, A.; Benussi, L.; Bianco, S.; Piccolo, D.; Primavera, F.; Cimmino, A.; Crucy, S.; Rios, A.A.O.; Tytgat, M.; Zaganidis, N.; Gul, M.; Fagot, A.; Bhatnagar, V.; Singh, J.; Kumari, R.; Mehta, A.; Ahmad, A.; Awan, I.M.; Shahzad, H.; Hoorani, H.; Asghar, M.I.; Muhammad, S.; Ahmed, W.; Shah, M.A.; Cho, S.W.; Choi, S.Y.; Hong, B.; Kang, M.H.; Lee, K.S.; Lim, J.H.; Park, S.K.; Kim, M.S.; Laktineh, I.B.; Lagarde, F.; Gouzevitch, M.; Grenier, G.; Pedraza, I.; Bernardino, S. Carpinteyro; Estrada, C. Uribe; Carrillo Moreno, S.; Valencia, F. Vazquez; Pant, L.M.; Buontempo, S.; Cavallo, N.; Fabozzi, F.; Orso, I.; Lista, L.; Meola, S.; Merola, M.; Paolucci, P.; Thyssen, F.; Lanza, G.; Esposito, M.; Braghieri, A.; Magnani, A.; Riccardi, C.; Salvini, P.; Vai, I.; Vitulo, P.; Montagna, P.; Ban, Y.; Qian, S.J.; Choi, M.; Choi, Y.; Goh, J.; Kim, D.; Dimitrov, A.; Litov, L.; Petkov, P.; Pavlov, B.; Bagaturia, I.; Lomidze, D.; Avila, C.; Cabrera, A.; Sanabria, J.C.; Crotty, I.; Vaitkus, J.

2016-01-01

The CMS experiment, located at the CERN Large Hadron Collider, has a redundant muon system composed by three different detector technologies: Cathode Strip Chambers (in the forward regions), Drift Tubes (in the central region) and Resistive Plate Chambers (both its central and forward regions). All three are used for muon reconstruction and triggering. During the first long shutdown (LS1) of the LHC (2013-2014) the CMS muon system has been upgraded with 144 newly installed RPCs on the forth forward stations. The new chambers ensure and enhance the muon trigger efficiency in the high luminosity conditions of the LHC Run2. The chambers have been successfully installed and commissioned. The system has been run successfully and experimental data has been collected and analyzed. The performance results of the newly installed RPCs will be presented.

Primary circuit water chemistry during shutdown period at Kalinin NPP

International Nuclear Information System (INIS)

Gorbatenko, S.; Otchenashev, G.; Yurmanov, V.

2005-01-01

The primary circuit water chemistry feature at Kalinin NPP is using of special up-dated regime during the period of unit shutdown for refueling. The main objective of up-dated regime is removing from the circuit long time living corrosion products on SVO-2 ion exchange filters with the purpose of dose rates reduction from the equipment and in such a way reduction of maintenance personnel overexposure. (N.T.)

77 FR 65119 - Approval and Promulgation of Implementation Plans; Texas; Revisions to the New Source Review (NSR...

Science.gov (United States)

2012-10-25

...), Continuous Parameter Monitoring System (CPMS), and Predictive Emissions Monitoring System (PEMS). [cir... record (and in some cases report) emissions events, which include unscheduled maintenance, startup, and... shutdown activities'' with ``emissions from planned maintenance, startup, or shutdown activities,'' which...

A Bayesian reliability study on motorized valves for the emergency core cooling, heat transport isolation and shutdown cooling systems at Gentilly-2 Nuclear Generating Station

International Nuclear Information System (INIS)

Smith, J.E.; Rennick, D.F.; Nainer, A.

1996-01-01

The objective of this is to examine operational data on 32 motorized valves in the emergency core cooling, shutdown cooling and heat transport isolation systems and determine if the evidence would support a reduction in testing frequency of these valves. The methodology used is to examine the data which has accumulated on motorized valve failures since Gentilly-2 first entered service, compare these data with similar data from other sources, and determine whether the evidence indicate that demand-based, wear out type failure mechanisms play a significant role in the recorded failures. The statistical data are then updated, using a Bayesian updating procedure, to obtain revised time based failure rates and demand based probabilities of failure on demand for the motorized valves. The revised failure rates and probabilities are then applied to the fault tree models for the systems of interest to determine what effects there would be, with the current test intervals and with extended test intervals, on the probability of failure of the systems. (author)

Corrosion product behaviour in the Loviisa nuclear power plant primary coolant: measures taken to lower radiation levels by modified shutdown procedures

International Nuclear Information System (INIS)

Jaernstroem, R.T.

1983-01-01

The primary circuit chemistry of the Loviisa nuclear power plant differs in some respects from the concepts commonly used in PWRs. In general, Loviisa 1, which is now in its sixth cycle, and Loviisa 2, which is in its second refuelling and maintenance shutdown (October 1982), are very clean compared with several other PWRs and it seems to be possible to keep the radiation levels low and even reduce them by using correct chemistry during operation; the shutdown conditions seem to have great influence on this matter. These modified shutdown conditions and their influence on radiation levels, dose rates and radwaste buildup are discussed. (author)

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Effect of the Online Game Shutdown Policy on Internet Use, Internet Addiction, and Sleeping Hours in Korean Adolescents.

Science.gov (United States)

Choi, Jiyun; Cho, Hyunseok; Lee, Seungmin; Kim, Juyeong; Park, Eun-Cheol

2018-05-01

Internet addiction has emerged as a major public health problem worldwide. In November 2011, the South Korean government implemented an online game shutdown policy, lasting from 12:00 to 6:00 am, as a means of preventing Internet addiction in adolescents aged 15 or below. This study analyzed the effect of this shutdown policy on adolescent Internet use, addiction, and sleeping hours. We analyzed data collected from the Korea Youth Risk Behavior Web-based Survey from 2011 to 2015. Respondents were divided into two groups by age: aged 15 or below (male = 76,048, female = 66,281) and aged 16 or above (male = 52,568, female = 49,060). A difference-in-difference analysis was used to evaluate the effect of this shutdown policy. In 2012, which is immediately following policy enforcement, daily amount of Internet use (in minutes) decreased more in adolescents affected by the policy (i.e., the aged 15 or below group). However, it steadily increased in 2013, 2014, 2015, and showed no meaningful long-term improvements 4 years after policy implementation (-3.648 minutes in 2012 [p = .001], -3.204 minutes in 2013 [p = .011], -1.140 minutes in 2014 [p = .384], and 2.190 minutes in 2015 [p = .107]). The shutdown policy did not alter Internet addiction or sleeping hours. Interestingly, female adolescents, adolescents with low academic performance, and adolescents with low exercise levels exhibited comparatively stronger and longer lasting initial declines in Internet usage. The shutdown policy had practically insignificant effects in reducing Internet use for target adolescents. Thus, policymakers aiming to reduce or prevent Internet addiction should use different strategies. Copyright © 2017 The Society for Adolescent Health and Medicine. Published by Elsevier Inc. All rights reserved.

Acquisition system of tandem injector parameters

International Nuclear Information System (INIS)

Decourt, M.

1986-01-01

The system centralizes all the parameters belonging to the accelerator injector. The acquisition center system reinforces an original device made of cameras and video receivers. Besides giving access to all the parameters of the ion source, the new system allows, in the ''OSCILLO'' mode, to visualize in real time any channel on the oscilloscope [fr

Transient fission product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.

1995-01-01

Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Allard, L.T.; Elerath, J.G.

1976-02-01

This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

CAREM-25. Purification and volume control system

International Nuclear Information System (INIS)

Acosta, Eduardo; Carlevaris, Rodolfo; Patrignani, Alberto; Chocron, Mauricio; Goya, Hector E.; Ortega, Daniel A.; Ramilo, Lucia B.

2000-01-01

The purification and volume control system has the following main functions: water level control inside reactor pressure vessel (RPR) in all the reactor operational modes, pressure control when the reactor operates in solid state, and maintenance of radiological, physical and chemical parameters of primary water. In case of Hot Shutdown operational mode and also after Scram the system is capable of extraction of nuclear decay heat. The design of the system is in accordance with the Requirements of ANSI/ ANS 51.1; 58.11 and 56.2 standards. (author)

Uncertainty reduction requirements in cores designed for passive reactivity shutdown

International Nuclear Information System (INIS)

Wade, D.C.

1988-01-01

The first purpose of this paper is to describe the changed focus of neutronics accuracy requirements existing in the current US advanced LMR development program where passive shutdown is a major design goal. The second purpose is to provide the background and rationale which supports the selection of a formal data fitting methodology as the means for the application of critical experiment measurements to meet these accuracy needs. 6 refs., 1 fig., 2 tabs

Risk-based evaluation of Allowed Outage Times (AOTs) considering risk of shutdown

International Nuclear Information System (INIS)

Mankamo, T.; Kim, I.S.; Samanta, P.K.

1992-01-01

When safety systems fail during power operation, Technical Specifications (TS) usually limit the repair within Allowed Outage Time (AOT). If the repair cannot be completed within the AOT, or no AOT is allowed, the plant is required to be shut down for the repair. However, if the capability to remove decay heat is degraded, shutting down the plant with the need to operate the affected decay-heat removal systems may impose a substantial risk compared to continued power operation over a usual repair time. Thus, defining a proper AOT in such situations can be considered as a risk-comparison between the repair in frill power state with a temporarily increased level of risk, and the altemative of shutting down the plant for the repair in zero power state with a specific associated risk. The methodology of the risk-comparison approach, with a due consideration of the shutdown risk, has been further developed and applied to the AOT considerations of residual heat removal and standby service water systems of a boiling water reactor (BWR) plant. Based on the completed work, several improvements to the TS requirements for the systems studied can be suggested

Magnetic latch trigger for inherent shutdown assembly

International Nuclear Information System (INIS)

Sowa, E.S.

1976-01-01

An inherent shutdown assembly for a nuclear reactor is provided. A neutron absorber is held ready to be inserted into the reactor core by a magnetic latch. The latch includes a magnet whose lines of force are linked by a yoke of material whose Curie point is at the critical temperature of the reactor at which the neutron absorber is to be inserted into the reactor core. The yoke is in contact with the core coolant or fissionable material so that when the coolant or the fissionable material increase in temperature above the Curie point the yoke loses its magnetic susceptibility and the magnetic link is broken, thereby causing the absorber to be released into the reactor core. 6 claims, 3 figures

Reactor protection and shut-down system

International Nuclear Information System (INIS)

Klar

1980-01-01

The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

Science.gov (United States)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

Shutdown channels and fitted interlocks in atomic reactors

International Nuclear Information System (INIS)

Furet, J.; Landauer, C.

1968-01-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

Cryogenics system: strategy to achieve nominal performance and reliable operation

International Nuclear Information System (INIS)

Bremer, J.; Brodzinski, K.; Casas, J.; Claudet, S.; Delikaris, D.; Delruelle, N.; Ferlin, G.; Fluder, C.; Perin, A.; Perinic, G.; Pezzetti, M.; Pirotte, O.; Tavian, L.; Wagner, U.

2012-01-01

During the LHC operation in 2010 and 2011, the cryogenic system has achieved an availability level fulfilling the overall requirement. To reach this level, the cryogenic system has profited like many other beam-dependent systems from the reduced beam parameters. Therefore, impacts of some failures occurred during the LHC operation were mitigated by using the over-capacity margin, the existing built-in redundancy in between adjacent sector cryogenic plants and the 'cannibalization' of spares on two idle cryogenic plants. These two first years of operation were also crucial to identify the weaknesses of the present cryogenic maintenance plan and new issues like SEUs. After the LS1, nominal beam parameters are expected and the mitigated measures will be less effective or not applicable at all. Consequently, a consolidation plan to improve the MTBF and the MTTR of the LHC cryogenic system is under definition. Concerning shutdown periods, the present cryogenic sectorization imposes some restrictions in the type of interventions (e.g. cryo-magnet removal) which can be done without affecting the operating conditions of the adjacent sector. This creates additional constrains and possible extra down-time in the schedule of the shutdowns including the hardware commissioning. This presentation focuses on the consolidation plan foreseen during the LS1 to improve the performance of the LHC cryogenic system in terms of availability and sectorization. (authors)

Exploitation examination of reliability of coal dust systems

International Nuclear Information System (INIS)

Dojchinovski, Ilija; Trajkovski, Kole

1997-01-01

Designers and operators wish is, long, failure free operation at designed parameters of every system. Always we know the system start up time, but we don't know how long this system will operate successfully. Because of that in this article is given a method how, step by step, to determine the reliability of the system. Reliability parameters are obtained from experimental and operational data. When reliability parameters are determined then it is very easy to compare reliability of similar systems, for example excavators, or different systems, such as truck and rubber band transport system. Practical use of the theory of reliability is by purchasing of the systems when manufacturers have to have and present reliability parameters and on this way we can decide which system satisfies our needs regarding the quality-price-reliability. Reliability can be practically used in system operation where: 1) system reliability is maintained with proper start, use and shutdown of the system; 2) a system reliability is maintained with good maintenance organization; 3) a system reliability is maintained with innovations and improvements with final purpose removing of the imperfections experienced through the operation. Reliability is very important parameter in power generation plants. (Author)

Optimization of startup and shutdown operation of simulated moving bed chromatographic processes.

Science.gov (United States)

Li, Suzhou; Kawajiri, Yoshiaki; Raisch, Jörg; Seidel-Morgenstern, Andreas

2011-06-24

This paper presents new multistage optimal startup and shutdown strategies for simulated moving bed (SMB) chromatographic processes. The proposed concept allows to adjust transient operating conditions stage-wise, and provides capability to improve transient performance and to fulfill product quality specifications simultaneously. A specially tailored decomposition algorithm is developed to ensure computational tractability of the resulting dynamic optimization problems. By examining the transient operation of a literature separation example characterized by nonlinear competitive isotherm, the feasibility of the solution approach is demonstrated, and the performance of the conventional and multistage optimal transient regimes is evaluated systematically. The quantitative results clearly show that the optimal operating policies not only allow to significantly reduce both duration of the transient phase and desorbent consumption, but also enable on-spec production even during startup and shutdown periods. With the aid of the developed transient procedures, short-term separation campaigns with small batch sizes can be performed more flexibly and efficiently by SMB chromatography. Copyright © 2011 Elsevier B.V. All rights reserved.

Parameter identification of chaos system based on unknown parameter observer

International Nuclear Information System (INIS)

Wang Shaoming; Luo Haigeng; Yue Chaoyuan; Liao Xiaoxin

2008-01-01

Parameter identification of chaos system based on unknown parameter observer is discussed generally. Based on the work of Guan et al. [X.P. Guan, H.P. Peng, L.X. Li, et al., Acta Phys. Sinica 50 (2001) 26], the design of unknown parameter observer is improved. The application of the improved approach is extended greatly. The works in some literatures [X.P. Guan, H.P. Peng, L.X. Li, et al., Acta Phys. Sinica 50 (2001) 26; J.H. Lue, S.C. Zhang, Phys. Lett. A 286 (2001) 148; X.Q. Wu, J.A. Lu, Chaos Solitons Fractals 18 (2003) 721; J. Liu, S.H. Chen, J. Xie, Chaos Solitons Fractals 19 (2004) 533] are only the special cases of our Corollaries 1 and 2. Some observers for Lue system and a new chaos system are designed to test our improved method, and simulations results demonstrate the effectiveness and feasibility of the improved approach

Reactor protection systems of 500 MWe PHWRs

Energy Technology Data Exchange (ETDEWEB)

Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

1997-03-01

The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

Development of a remote monitoring system, through monitoring of key safety parameters for a nuclear research reactor

International Nuclear Information System (INIS)

Urcia, Agustin; Arrieta, Rolando; Baltuano, Oscar; Chan, Renzo; Tincopa, Jean Pierre; Urquizo, Rafael; Rosas, Bernick

2014-01-01

This paper presents the detailed development, installation and commissioning of water level sensors and exposure rate range in the 11 meters level (mouth of tank) of the RP-10 nuclear reactor used to continuously monitor these values and use them as security for the periods of no presence of operating personnel (overlooking situation) with the reactor in shutdown state. The levels of these parameters are packaged and transmitted to a controller in the control room of reactor for display and activation of alarm levels. Additionally, the design of these warning signs is shown in conjunction with the fire alarm in the building of reactor and auxiliary laboratories to be transmitted to the physical security facility, located at a distance of 500 meters. (authors).

Behaviour of ruthenium in the case of shutdown of the cooling system of HLLW storage tanks

International Nuclear Information System (INIS)

Philippe, M.; Gue, J.P.; Mercier, J.P.

1990-12-01

The consequences of the failure of the cooling system of fission product storage tanks over a variable period were investigated as part of the safety analysis of the La Hague spent fuel reprocessing plant. Due to the considerable heat release, induced by the fission products, a prolonged shutdown of the tank cooling system could cause the progressive evaporation of the solutions to dryness, and culminate in the formation of volatile species of ruthenium and their release in the tank venting circuit. To determine the fraction of ruthenium likely to be transferred from the storage tanks in volatile or aerosol form during the failure, evaporation tests were conducted by evaporating samples of actual nitric acid solutions of fission products, obtained on the laboratory scale after the reprocessing of several kilograms of MOX fuels irradiated to 30.000 MW day ·t -1 . A distillation apparatus was designed to operate with small-volume solution samples, reproducing the heating conditions existing in the reprocessing plant within a storage tank for fission products. The main conclusions drawn from these experiments are as follows: - ruthenium is only volatilized in the final phase of evaporation, just before desiccation, - for a final temperature limited to 160 deg. C, the total fraction of volatilized ruthenium reaches 12%, - in the presence of H 2 O, HNO 3 , NO x and O 2 , the volatilized ruthenium recombines mainly in the form of ruthenium nitrosyl nitrates, or decomposes into ruthenium oxide (probably RuO 2 ) on the walls of the apparatus. Assuming a heating power density of 10 W/liter of concentrate, and a perfectly adiabatic storage system, the minimum time required to reach dryness can be estimated at 90 h, allowing substantial time to take action to restore a cooling source. It is probable that, in an industrial storage tank, the heat losses from the tank and the offgas discharge ducts will cause recondensation and internal reflux, which will commensurately delay

Dynamic Response of AP1000 Nuclear Island Due to Safe Shutdown Earthquake Loading

Directory of Open Access Journals (Sweden)

Gan Buntara S.

2017-01-01

Full Text Available AP1000 is a standard nuclear power plant developed by Westinghouse and its partners by using an advanced passive safety feature. Among the five principle building structures, namely the nuclear island, turbine building, annex building, diesel generator building and radwaste building, the safety of the nuclear island building is the most concerned. This paper investigates the dynamic response of the nuclear island building of the AP1000 plant subjected to safe shutdown earthquake loadings. A finite element model for the building, which is assumed to be built in a hard-rock base, is developed and its dynamic response is computed with the aid of the commercial finite element package ANSYS. The dynamic characteristics, including the natural frequencies, the vibration modes, and the time histories for displacements, velocities, and accelerations of the building are obtained for two typical safe shutdown earthquakes, El Centro and Kobe earthquakes. The dynamic behavior of the building due to the earthquakes and its safety is examined and highlighted.

Best estimate analysis of the thermal expansion scenario during shutdown in a PWR

International Nuclear Information System (INIS)

Macian, R.; Nechvatal, L.

2001-01-01

In this paper we examine the consequences following the hypothetical failure of the Residual Heat Removal (RHR) system during the shutdown operating mode in a Pressurized Water Reactor (PWR). If the RHR system decay heat removal capability cannot be ensured, then the decay heat released in the core will heat up the Reactor Coolant System (RCS) inventory and will cause it to expand. If the thermal expansion is such that the entire RCS becomes ''water-solid'', that is, completely filled with water, then further expansion will result in a rapid increase of the RCS pressure. Such a situation could threaten the integrity of the RCS pressure boundary and lead to a dangerous break in the primary system or in the lines of the systems connected to it, e.g. RHR system. The pressure increase can be arrested by the opening of the pressurizer relief valves (PORVs) or, in those PWRs in which the RHR system is not isolated after it fails, by the opening of the pressure relief valve in the RHR system line. The purpose of the analyses presented in this paper is to determine whether mitigating measures, such as the opening of only one of the PORV and the RHR relief valve, are capable of preventing a fast pressure increase. (author)

Containment closure time following loss of cooling under shutdown conditions of YGN units 3 and 4

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Kim, Se Won; Kim, Hho Jung

1998-01-01

The YGN Units 3 and 4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling. The thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior. From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. These data provide useful information to the abnormal procedure to cope with the event

A loss-of -RHR event under the various plant configurations in low power or shutdown conditions

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Lee, Suk Ho; Kim, Hho Jung

1997-01-01

A present study addresses a loss-of-RHR event as an initiating event under specific low power of shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region

40 CFR 60.2120 - What happens during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 6 2010-07-01 2010-07-01 false What happens during periods of startup... 1, 2001 Emission Limitations and Operating Limits § 60.2120 What happens during periods of startup... during CISWI unit startups, shutdowns, or malfunctions. (b) Each malfunction must last no longer than 3...

CORAL and COOL during the LHC long shutdown.

CERN Document Server

Valassi, Andrea; Dulstra, D; Goyal, N; Salnikov, A; Trentadue, R; Wache, M

2014-01-01

CORAL and COOL are two software packages used by the LHC experiments for managing detector conditions and other types of data using relational database technologies. They have been developed and maintained within the LCG Persistency Framework, a common project of the CERN IT department with ATLAS, CMS and LHCb. This presentation reports on the status of CORAL and COOL at the time of CHEP2013, covering the new features and enhancements in both packages, as well as the changes and improvements in the software process infrastructure. It also reviews the usage of the software in the experiments and the outlook for ongoing and future activities during the LHC long shutdown (LS1) and beyond.

CORAL and COOL during the LHC long shutdown

CERN Multimedia

Valassi, A; Dykstra, D; Goyal, N; Salnikov, A; Trentadue, R; Wache, M

2013-01-01

CORAL and COOL are two software packages used by the LHC experiments for managing detector conditions and other types of data using relational database technologies. They have been developed and maintained within the LCG Persistency Framework, a common project of the CERN IT department with ATLAS, CMS and LHCb. This presentation reports on the status of CORAL and COOL at the time of CHEP2013, covering the new features and enhancements in both packages, as well as the changes and improvements in the software process infrastructure. It also reviews the usage of the software in the experiments and the outlook for ongoing and future activities during the LHC long shutdown (LS1) and beyond.

Containment closure time following the loss of shutdown cooling event of YGN Units 3 and 4

International Nuclear Information System (INIS)

Seul, Kwang Won; Bang, Young Seok; Kim, Hho Jung

1999-01-01

The YGN Units 3 and 4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOS3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide using information to operators to cope with the loss of SDC event. (Author). 15 refs., 3 tabs., 7 figs

Methods for nondestructive assay holdup measurements in shutdown uranium enrichment facilities

International Nuclear Information System (INIS)

Hagenauer, R.C.; Mayer, R.L. II.

1991-09-01

Measurement surveys of uranium holdup using nondestructive assay (NDA) techniques are being conducted for shutdown gaseous diffusion facilities at the Oak Ridge K-25 Site (formerly the Oak Ridge Gaseous Diffusion Plant). When in operation, these facilities processed UF 6 with enrichments ranging from 0.2 to 93 wt % 235 U. Following final shutdown of all process facilities, NDA surveys were initiated to provide process holdup data for the planning and implementation of decontamination and decommissioning activities. A three-step process is used to locate and quantify deposits: (1) high-resolution gamma-ray measurements are performed to generally define the relative abundances of radioisotopes present, (2) sizable deposits are identified using gamma-ray scanning methods, and (3) the deposits are quantified using neutron measurement methods. Following initial quantitative measurements, deposit sizes are calculated; high-resolution gamma-ray measurements are then performed on the items containing large deposits. The quantitative estimates for the large deposits are refined on the basis of these measurements. Facility management is using the results of the survey to support a variety of activities including isolation and removal of large deposits; performing health, safety, and environmental analyses; and improving facility nuclear material control and accountability records. 3 refs., 1 tab

Government Shutdown: Operations of Department of Defense During a Lapse in Appropriations

Science.gov (United States)

2011-04-01

proactive in working with creditors to reschedule debt repayments under these circumstances… c. Military personnel: During a shutdown of DoD activities due...creditors to reschedule debt repayments under these circumstances. The key point that both the creditor and the soldier should remember is that the...including Uniformed Services Treatment Facilities) including doctors, nurses , medical technicians, dentists, and essential support personnel (cooks

Reactivity changes in hybrid thermal-fast reactor systems during fast core flooding

International Nuclear Information System (INIS)

Pesic, M.

1994-09-01

A new space-dependent kinetic model in adiabatic approximation with local feedback reactivity parameters for reactivity determination in the coupled systems is proposed in this thesis. It is applied in the accident calculation of the 'HERBE' fast-thermal reactor system and compared to usual point kinetics model with core-averaged parameters. Advantages of the new model - more realistic picture of the reactor kinetics and dynamics during local large reactivity perturbation, under the same heat transfer conditions, are underlined. Calculated reactivity parameters of the new model are verified in the experiments performed at the 'HERBE' coupled core. The model has shown that the 'HERBE' safety system can shutdown reactor safely and fast even in the case of highly set power trip and even under conditions of big partial failure of the reactor safety system (author)

Experience in ultrasonic gap measurement between calandria tubes and liquid injection shutdown systems nozzles in Bruce Nuclear Generating Station

International Nuclear Information System (INIS)

Abucay, R.C.; Mahil, K.S.; Goszczynski, J.J.

1995-01-01

The gaps between calandria tubes (CT) and Liquid Injection Shutdown System (LISS) nozzles at the Bruce Nuclear Generating Station ''A'' (Bruce A) are known to decrease with time due to radiation induced creep/sag of the calandria tubes. If this gap decreases to a point where the calandria tubes come into contact with the LISS nozzle, the calandria tubes could fail as a result of fretting damage. Proximity measurements were needed to verify the analytical models and ensure that CT/LISS nozzle contact does not occur earlier than predicted. The technique used was originally developed at Ontario Hydro Technologies (formerly Ontario Hydro Research Division) in the late seventies and put into practical use by Research and Productivity Council (RPC) of New Brunswick, who carried out similar measurements at Point Lepreau NGS in 1989 and 1991. The gap measurement was accomplished y inserting an inspection probe, containing four ultrasonic transducers (2 to measure gaps and 2 to check for probe tilt) and a Fredericks electrolytic potentiometer as a probe rotational sensor, inside LISS Nozzle number-sign 7. The ultrasonic measurements were fed to a system computer that was programmed to convert the readings into fully compensated gaps, taking into account moderator heavy water temperature and probe tilt. Since the measured gaps were found to be generally larger than predicted, the time to CT/LISS nozzle contact is now being re-evaluated and the planned LISS nozzle replacement will likely be deferred, resulting in considerable savings

Action plan during reactor shutdown in October 1965, Annex 5; Prilog br. 5 - Plan radova u toku stajanja reaktora u mesecu oktobru 1965. godine

Energy Technology Data Exchange (ETDEWEB)

Nikolic, M [Reaktor RA, Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1965-12-15

The action plan of the division for reactor maintenance during reactor shutdown includes detailed list of tasks for mechanics, electronic and electrical equipment group during the reactor shutdown period in October 1965. It contains tasks for planned shutdown periods in September, August, July, May, April, March, and February 1965. [Serbo-Croat] Plan radova Odelenja odrzavanja reaktora RA za period stajanja reaktora u oktobru mesecu 1965. sadrzi detaljnu listu zadataka masinske grupe, elektro grupe i elektronske grupe. Ovaj prilog sadrzi i zadatke koji ce biti obavljeni tokom planiranih perioda kada je reaktor zaustavljen u septembru, avgustu, julu, junu, maju, aprilu, martu i februaru 1965.

A new concept of safety parameter display system

International Nuclear Information System (INIS)

Martinez, A.S.; Oliveira, L.F.S. de; Schirru, R.; Thome Filho, Z.D.; Silva, R.A. da.

1986-07-01

A general description of Angra-1 Parameter Display System (SSPA), a real time and on-line computerized monitoring system for the parameters related to the power plant safety is presented. This system has the main purpose of diminish the load on the Angra-1 power plant operators at an emergency event by supplying them with the additional tools serving as the basis for a prompt identification of the accident. The SSPA is a kind of safety parameter display system whose concept was introduced after Three Mile Island accident in USA. The SSPA comprises two nuclear applications independently considered. They are included into the Parameters Monitoring Integrated System (SIMP) and the safety critical function system (SFCS). (Author) [pt

Firefighter safety for PV systems: Overview of future requirements and protection systems

DEFF Research Database (Denmark)

Spataru, Sergiu; Sera, Dezso; Blaabjerg, Frede

2013-01-01

for operators during maintenance or fire-fighting. One of the solutions is individual module shutdown by short-circuiting or disconnecting each PV module from the PV string. However, currently no standards have been adopted either for implementing or testing these methods, or doing an evaluation of the module...... shutdown procedures. This paper gives an overview on the most recent fire - and firefighter safety requirements for PV systems, with focus on system and module shutdown systems. Several solutions are presented, analyzed and compared by considering a number of essential characteristics, including......An important and highly discussed safety issue for photovoltaic systems is that, as long as they are illuminated, a high voltage is present at the PV string terminals and cables between the string and inverters, independent of the state of the inverter's dc disconnection switch, which poses a risk...

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

International Nuclear Information System (INIS)

Biondo, Elliott D.; Wilson, Paul P. H.

2017-01-01

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 _± 5 • _1_0_"_4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

Parameter identifiability of linear dynamical systems

Science.gov (United States)

Glover, K.; Willems, J. C.

1974-01-01

It is assumed that the system matrices of a stationary linear dynamical system were parametrized by a set of unknown parameters. The question considered here is, when can such a set of unknown parameters be identified from the observed data? Conditions for the local identifiability of a parametrization are derived in three situations: (1) when input/output observations are made, (2) when there exists an unknown feedback matrix in the system and (3) when the system is assumed to be driven by white noise and only output observations are made. Also a sufficient condition for global identifiability is derived.

A methodology for Level 2 PSA evaluation with consideration of specific features for Low Power Shutdown Probabilistic Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Gab; Seok, Ho [KEPCO-ENC, Yongin (Korea, Republic of)

2015-05-15

The primary objective of the Level 2 PSA during Lower Power/Shutdown (LPSD) operation is to provide insights into potential plant vulnerabilities with regard to accident progression. The shutdown risk information can be used to provide the information to develop outage risk management guidelines. The LPSD Level 2 analysis utilizes much of the at-power Level 2 analysis for bounding, conservative treatment of severe accident phenomena. But, for some portions of the analysis including Plant Operational States (POSs), LPSD-specific evaluations such as UPC related to the containment Equipment Hatch (E/H) with 4 bolts, Reactor Coolant System (RCS) Not Intact for severe accident phenomena are desired for realistic evaluation. All POSs are evaluated for their Large Release Frequency (LRF). Some POSs are evaluated conservatively utilizing the at-power models, and other POSs are evaluated in specific analysis. The overall LPSD Level 2 model is evaluated. If the containment E/H and one of the two doors on each of the personal air locks are closed as containment is operable at reduced RCS inventory operation, LRF is expected to be less than 10% of LPSD CDF.

A scoping evaluation of severe accidents at Surry and Grand Gulf Nuclear Power Plants resulting from earthquakes during shutdown conditions

International Nuclear Information System (INIS)

Budnitz, R.J.; Davis, P.R.

1991-01-01

This report explores the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions at two nuclear power plants, Surry Unit I and Grand Gulf Unit 1. The effort is scoping in character, and has been performed primarily to establish if a potential problem exists sufficient to justify a more rigorous and more quantitative evaluation. A summary is presented of the important conclusions that have been reached. The most important conclusion is that the core-damage frequencies for earthquake-initiated accidents during shutdown at both Surry Unit I and Grand Gulf Unit I are found to be low in absolute terms. The reasons for this are that in their ability to respond to earthquakes during shutdowns, the plants both have large seismic capacities, well above their design-basis levels; and also that both sites enjoy among the lowest seismic hazards of any LWR sites in the US

Containment closure time following loss of cooling under shutdown conditions of YGN units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Seul, Kwang Won; Bang, Young Seok; Kim, Se Won; Kim, Hho Jung [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

1998-12-31

The YGN Units 3 and 4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling. The thermal hydraulic analyses were performed for the five cases of RCS configurations under the worst event scenario, unavailable secondary cooling and no RCS inventory makeup, using the RELAP5/MOD3.2 code to investigate the plant behavior. From the analyses results, times to boil, times to core uncovery and times to core heat up were estimated to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. These data provide useful information to the abnormal procedure to cope with the event. 6 refs., 7 figs., 2 tabs. (Author)

Study on the identification of main drivers affecting the performance of human operators during low power and shutdown operation

International Nuclear Information System (INIS)

Kim, Ar Ryum; Park, Jinkyun; Kim, Ji Tae; Kim, Jaewhan; Seong, Poong Hyun

2016-01-01

Highlights: • The performance of human operator during LPSD operation is significantly important. • Human performance is affected by drivers such as procedure, training, and etc. • Main drivers during LPSD operation at domestic NPPs were suggested. • It is expected that it will be used for estimating human reliability during LPSD operation. - Abstract: In the past, many researchers believed that a reactor during low power and shutdown operation was sufficiently safe. This belief has been changed by the number of accidents during such types of operation, which is significantly high. Also, it was pointed out that one of the main differences between low power and shutdown operation and full power operation is the significance of human action because there are huge amounts of human actions due to extensive maintenance and testing while automatic control and safety functions may be disabled and procedures are insufficient or incomplete. This paper suggests the main drivers in performing human reliability analysis. For this study, we reviewed eight reports relating to human performance during low power and shutdown operation and applied a root cause analysis method for 53 human or human-related events at domestic nuclear power plants to derive the main drivers that affect the occurrence of those events. As a result, several main drivers were derived, such as procedures, training, experience of personnel, and workload/stress. It is expected that these main drivers will be used to perform human reliability analysis for low power and shutdown operation.

Shutdown dose rates at ITER equatorial ports considering radiation cross-talk from torus cryopump lower port

Energy Technology Data Exchange (ETDEWEB)

Juárez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Pampin, Raul [F4E, Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Levesy, Bruno [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Moro, Fabio [ENEA, Via Enrico Fermi 45, Frascati, Rome (Italy); Suarez, Alejandro [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

2015-11-15

Shutdown dose rates for planned maintenance purposes is an active research field in ITER. In this work the radiation (neutron and gamma) cross-talk between ports in the most conservative case foreseen in ITER is investigated: the presence of a torus cryopump lower port, mostly empty for pumping efficiency reasons. There will be six of those ports: #4, #6, #10, #12, #16 and #18. The equatorial ports placed above them will receive a significant amount of additional radiation affecting the shutdown dose rates during in situ maintenance activities inside the cryostat, and particularly in the port interspace area. In this study a general situation to all the equatorial ports placed above torus cryopump lower ports is considered: a generic diagnostics equatorial port placed above the torus cryopump lower port (LP#4). In terms of shutdown dose rates at equatorial port interspace after 10{sup 6} s of cooling time, 405 μSv/h has been obtained, of which 160 μSv/h (40%) are exclusively due to radiation cross-talk from a torus cryopump lower port. Equatorial port activation due to only “local neutrons” contributes 166 μSv/h at port interspace, showing that radiation cross-talk from such a lower port is a phenomenon comparable in magnitude to the neutron leakage though the equatorial port plug.

Optimization of Parameters of Asymptotically Stable Systems

Directory of Open Access Journals (Sweden)

Anna Guerman

2011-01-01

Full Text Available This work deals with numerical methods of parameter optimization for asymptotically stable systems. We formulate a special mathematical programming problem that allows us to determine optimal parameters of a stabilizer. This problem involves solutions to a differential equation. We show how to chose the mesh in order to obtain discrete problem guaranteeing the necessary accuracy. The developed methodology is illustrated by an example concerning optimization of parameters for a satellite stabilization system.

Diesel fuel filtration system

International Nuclear Information System (INIS)

Schneider, D.

1996-01-01

The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

A Comparative Study of Distribution System Parameter Estimation Methods

Energy Technology Data Exchange (ETDEWEB)

Sun, Yannan; Williams, Tess L.; Gourisetti, Sri Nikhil Gup

2016-07-17

In this paper, we compare two parameter estimation methods for distribution systems: residual sensitivity analysis and state-vector augmentation with a Kalman filter. These two methods were originally proposed for transmission systems, and are still the most commonly used methods for parameter estimation. Distribution systems have much lower measurement redundancy than transmission systems. Therefore, estimating parameters is much more difficult. To increase the robustness of parameter estimation, the two methods are applied with combined measurement snapshots (measurement sets taken at different points in time), so that the redundancy for computing the parameter values is increased. The advantages and disadvantages of both methods are discussed. The results of this paper show that state-vector augmentation is a better approach for parameter estimation in distribution systems. Simulation studies are done on a modified version of IEEE 13-Node Test Feeder with varying levels of measurement noise and non-zero error in the other system model parameters.

On-line use of personal computers to monitor and evaluate important parameters in the research reactor DHRUVA

International Nuclear Information System (INIS)

Sharma, S.K.; Sengupta, S.N.; Darbhe, M.D.; Agarwal, S.K.

1998-01-01

The on-line use of Personal Computers in research reactors, with custom made applications for aiding the operators in analysing plant conditions under normal and abnormal situations, has become extremely popular. A system has been developed to monitor and evaluate important parameters for the research reactor DHRUVA, a 100 MW research reactor located at the Bhabha Atomic Research Centre, Trombay. The system was essentially designed for on-line computation of the following parameters: reactor thermal power, reactivity load due to Xenon, core reactivity balance and performance monitoring of shut-down devices. Apart from the on-line applications, the system has also been developed to cater some off-line applications with Local Area Network in the Dhruva complex. The microprocessor based system is designed to function as an independent unit, parallel dumping the acquired data to a PC for application programmes. The user interface on the personal computer is menu driven application software written in 'C' language. The main input parameters required for carrying out the options given in the above menu are: Reactor power, Moderator level, Coolant inlet temperature to the core, Secondary coolant flow rate, temperature rise of secondary coolant across the heat exchangers, heavy water level in the Dump tank and Drop time of individual shut off rods. (author)

78 FR 38001 - Reconsideration of Certain Startup/Shutdown Issues: National Emission Standards for Hazardous Air...

Science.gov (United States)

2013-06-25

..., FRL-9827-1] RIN 2060-AR62 Reconsideration of Certain Startup/Shutdown Issues: National Emission... published in the Federal Register the proposed rule, ``Reconsideration of Certain New Source and Startup....'' That proposal opened for reconsideration certain issues, including those related to startup and...

Design, construction, operation, shutdown and surveillance of repositories for solid radioactive wastes in shallow ground

International Nuclear Information System (INIS)

1984-01-01

This report is a part of the IAEA publications under its Programme on Underground Disposal of Radioactive Wastes and is addressed to administrative and technical authorities and specialists who consider the shallow-ground disposal of low- and intermediate-level solid radioactive wastes of short half-lives. The report emphasizes the technological aspects, however it briefly discusses the safety philosophy and regulatory considerations too. The design, construction, operation, shutdown and surveillance of the repositories in shallow ground are considered in some detail, paying special attention to their interrelated aspects. In particular, a review is given of the following aspects: main design and construction considerations in relation to the natural features of the site; design and construction aspects during the repository development process; activities related to operational and post-operational stages of the repository; major steps in repository operation and essential activities in shutdown and operational and post-operational surveillance

A control system for industrial plant (e.g. a pressurized water reactor)

International Nuclear Information System (INIS)

Spiller, C.R.L.

1990-01-01

A control system for an industrial plant eg. a pressurised water nuclear reactor, comprises a plurality of instrument sets and a plurality of logic sets. The instrument sets have a number of sensors which detect parameters (temperature, pressure vibration) of the industrial plant, and have two serial link controllers which supply the output signals from each sensor in the instrument set sequentially to the logic sets via conductors. The logic sets have a number of auto select logic circuits, each of which selects data from the sensors from one of the instrument sets, and a synchroniser ensures that the output signals from the sensors detecting the same parameter are supplied to a voting logic circuit at the same time. The voting logic circuit performs a voting function on the output signals to produce a series of high reliability signals which are converted to parallel high reliability signals by a series to a parallel converter. The high reliability signals are supplied to a fault logic shutdown circuit which controls the operation of shutdown mechanisms for the industrial plant. (author)

Safe shutdown of Defense Program facilities at the Mound Plant, Miamisburg, Ohio

International Nuclear Information System (INIS)

Anderson, H.F.; Bantz, P.D.; Luthy, D.F.

1996-01-01

The Mound Plant was one of several production sites in the US Department of Energy's Defense Programs (DP) Weapons Complex. As a result of the downsizing of the weapons program, certain operations at Mound are being transferred to other DOE sites and the DP buildings at Mound are being shutdown. The objectives of the program are to reduce the hazardous and financial liabilities to DOE and to foster the reuse of facilities for economic development. The overall program is described. The process began with the categorization of excess DP buildings into three groups depending on their anticipated future use. The draft DOE/EM-60 Acceptance Criteria were used to develop a detailed shutdown checklist as the foundation of the process. The overall program budget, schedule, ad options for disposition of materials and components is presented. Accomplishments in FY94 and FY95 are described. By the end of FY95, all excess energetic materials and components, all excess chemicals (from non-radiation areas) and significant amounts of radioactive materials have been removed from the site. By the end of FY95, 47 of the 72 buildings in the program have been taken through all ten of the draft EM-60 acceptance criteria. Lessons learned, based on experience at Mound to date, are summarized

Spatio-temporal modeling of nonlinear distributed parameter systems

CERN Document Server

Li, Han-Xiong

2011-01-01

The purpose of this volume is to provide a brief review of the previous work on model reduction and identifi cation of distributed parameter systems (DPS), and develop new spatio-temporal models and their relevant identifi cation approaches. In this book, a systematic overview and classifi cation on the modeling of DPS is presented fi rst, which includes model reduction, parameter estimation and system identifi cation. Next, a class of block-oriented nonlinear systems in traditional lumped parameter systems (LPS) is extended to DPS, which results in the spatio-temporal Wiener and Hammerstein s

Complement of existing ASAMPSA2 guidance for Level 2 PSA for shutdown states of reactors, Spent Fuel Pool and recent R and D results

International Nuclear Information System (INIS)

Kumar, M.; Olsson, A.; Loeffler, H.; Morandi, S.; Gumenyuk, D.; Dejardin, P.; Yu, S.; Jan, P.; Kubicek, J.; Serrano, C.; Raimond, E.; Dirksen, G.; Ivanov, I.; Groudev, P.; Kowal, K.; Prosek, Andrej; Nitoi, M.; Vitazkova, J.; Hirata, K.; Burgazzi, L.

2016-01-01

This report can be considered as an addendum to the existing ASAMPSA2 guidance for Level 2 PSA. It provides complementary guidance for Level 2 PSA for accident in the NPP shutdown states and on spent fuel pool and comments on the importance of these accidents on nuclear safety. It includes also information on recent research and development useful for Level 2 PSA developments. The conclusions of the ASAMPSA-E end-users survey and of technical meetings of WP10, WP21, WP22, and WP30 at Vienna University in September 2014 which are relevant for Level 2 PSA have been reflected and are taken into account as much as it is possible with the current status of knowledge. For Level 2 PSA in shutdown states, two plant conditions are to be distinguished: - accident sequences with RPV head closed, - accident sequences with RPV head open. When the RPV head is closed, core melt accident phenomena are very similar to the sequences going on in full power mode. Therefore, the large body of guidance which is available for full power mode is basically applicable to shutdown mode with RPV closed as well. When the RPV is open, some of the L2 PSA issues become irrelevant compared to full power mode, while others come into existence. The situation is different for aspects which do not exist or which are less pronounced in sequences with RPV closed. The report also covers containment issues in shutdown states and discusses the applicability of existing guidance, potential gaps and deficiencies and recommendations are provided. For spent fuel pool accidents in Level 2 PSA, a set of issues is identified and addressed. If the spent fuel pool is located inside the containment, the potential release paths to the environment are almost the same as for core melt accidents in the RPV. If the spent fuel pool is located outside the containment, the potential release paths to the environment depend very much on plant specific properties, e.g. ventilation systems, building doors, roof under thermal

COMPUTING SERVICES DURING THE ANNUAL CERN SHUTDOWN

CERN Multimedia

2000-01-01

As in previous years, computing services run by IT division will be left running unattended during the annual shutdown. The following points should be noted. No interruptions are scheduled for local and wide area networking and the ACB, e-mail and unix interactive services. Maintenance work is scheduled for the NICE home directory servers and the central Web servers. Users must, therefore, expect service interruptions. Unix batch services will be available but without access to HPSS or to manually mounted tapes. Dedicated Engineering services, general purpose database services and the Helpdesk will be closed during this period. An operator service will be maintained and can be reached at extension 75011 or by email to: computer.operations@cern.ch Users should be aware that, except where there are special arrangements, any major problems that develop during this period will most likely be resolved only after CERN has reopened. In particular, we cannot guarantee backups for Home Directory files for eithe...

COMPUTING SERVICES DURING THE ANNUAL CERN SHUTDOWN

CERN Multimedia

2001-01-01

As in previous years, computing services run by IT division will be left running unattended during the annual shutdown. The following points should be noted. No interruptions are scheduled for local and wide area networking and the ACB, e-mail and unix interactive services. Unix batch services will be available but without access to manually mounted tapes. Dedicated Engineering services, general purpose database services and the Helpdesk will be closed during this period. An operator service will be maintained and can be reached at extension 75011 or by Email to computer.operations@cern.ch. Users should be aware that, except where there are special arrangements, any major problems that develop during this period will most likely be resolved only after CERN has reopened. In particular, we cannot guarantee backups for Home Directory files (for Unix or Windows) or for email folders. Any changes that you make to your files during this period may be lost in the event of a disk failure. Please note that all t...

Summary of Information Presented at an NRC-Sponsored Low-Power Shutdown Public Workshop, April 27, 1999, Rockville, Maryland

International Nuclear Information System (INIS)

Wheeler, Timothy A.; Whitehead, Donnie W.; Lois, Erasmia

1999-01-01

This report summarizes a public workshop that was held on April 27, 1999, in Rockville, Maryland. The workshop was conducted as part of the US Nuclear Regulatory Commission's (NRC) efforts to further develop its understanding of the risks associated with low power and shutdown operations at US nuclear power plants. A sufficient understanding of such risks is required to support decision-making for risk-informed regulation, in particular Regulatory Guide 1.174, and the development of a consensus standard. During the workshop the NRC staff discussed and requested feedback from the public (including representatives of the nuclear industry, state governments, consultants, private industry, and the media) on the risk associated with low-power and shutdown operations

Summary of Information Presented at an NRC-Sponsored Low-Power Shutdown Public Workshop, April 27, 1999, Rockville, Maryland

Energy Technology Data Exchange (ETDEWEB)

Wheeler, Timothy A.; Whitehead, Donnie W.; Lois, Erasmia

1999-07-01

This report summarizes a public workshop that was held on April 27, 1999, in Rockville, Maryland. The workshop was conducted as part of the US Nuclear Regulatory Commission's (NRC) efforts to further develop its understanding of the risks associated with low power and shutdown operations at US nuclear power plants. A sufficient understanding of such risks is required to support decision-making for risk-informed regulation, in particular Regulatory Guide 1.174, and the development of a consensus standard. During the workshop the NRC staff discussed and requested feedback from the public (including representatives of the nuclear industry, state governments, consultants, private industry, and the media) on the risk associated with low-power and shutdown operations.

Event data collection and database development during plant shutdown and low power operations at domestic and foreign reactors

International Nuclear Information System (INIS)

Kim, T. Y.; Park, J. H.; Han, S. J.; Im, H. K.; Jang, S. C.

2003-01-01

To reduce conservatism and to obtain completeness for Low Power and ShutDown(LPSD) PSA of nuclear plants, total of 625 event data have collected during shutdown and low power operations which have occurred during about 30 years at nuclear power plants of USA and European countries including 2 domestic events. To utilize efficiently these event data, a database program which is called LEDB (Low power and shutdown Event Database) was developed and all the event data collected were inserted in that program. By reviewing and analyzing these event data various way, a lot of very useful insights and ideas for preventing similar events from reoccurrence in domestic nuclear power plants can be obtained

The management of large cabling campaigns during the Long Shutdown 1 of LHC

CERN Document Server

Meroli, Stefano; Formenti, Fabio; Frans, Marten; Guillaume, Jean Claude; Ricci, Daniel

2014-01-01

The Large Hadron Collider at CERN entered into its first 18 month-long shutdown period in February 2013. During this period the entire CERN accelerator complex will undergo major consolidation and upgrade works, preparing the machines for LHC operation at nominal energy (7 TeV/beam). One of the most challenging activities concerns the cabling infrastructure (copper and optical fibre cables) serving the CERN data acquisition, networking and control systems. About 1000 kilometres of cables, distributed in different machine areas, will be installed, representing an investment of about 15 MCHF. This implies an extraordinary challenge in terms of project management, including resource and activity planning, work execution and quality control. The preparation phase of this project started well before its implementation, by defining technical solutions and setting financial plans for staff recruitment and material supply. Enhanced task coordination was further implemented by deploying selected competences to form a ...

Design Options for Thermal Shutdown Circuitry with Hysteresis Width Independent on the Activation Temperature

Directory of Open Access Journals (Sweden)

PLESA, C.-S.

2017-02-01

Full Text Available This paper presents several design options for implementing a thermal shutdown circuit with hysteretic characteristic, which has two special features: a programmable activation temperature (the upper trip point of the characteristic and a hysteresis width largely insensitive to the actual value of the activation temperature and to variations of the supply voltage. A fairly straightforward architecture is employed, with the hysteresis implemented by a current source enabled by the output of the circuit. Four possible designs are considered for this current source: VBE/R, modified-VBE/R, Widlar and a peaking current source tailored for this circuit. First, a detailed analytical analysis of the circuit implemented with these current sources is performed; it indicates the one best suited for this application and provides key sizing equations. Next, the chosen current source is employed to design the thermal shutdown protection of an integrated Low-Dropout Voltage Regulator (LDO for automotive applications. Simulation results and measurements performed on the silicon implementation fully validate the design. Moreover, they compare favorably with the performance of similar circuits reported recently.

Development of a safety and regulation systems simulation program II

International Nuclear Information System (INIS)

1985-05-01

This report describes the development of a safety and regulation systems simulation program under contract to the Atomic Energy Control Board of Canada. A systems logic interaction simulation (SLISIM) program was developed for the AECB's HP-1000 computer which operates in the interactive simulation (INSIM) program environment. The SLISIM program simulates the spatial neutron dynamics, the regulation of the reactor power and in this version the CANDU-PHW 600 MW(e) computerized shutdown systems' trip parameters. The modular concept and interactive capability of the INSIM environment provides the user with considerable flexibility of the setup and control of the simulation

WIMS-AECL/RFSP code validation of reactivity calculations following a long shutdown using the simple-cell history-based method

International Nuclear Information System (INIS)

Ardeshiri, F.; Donnelly, J.V.; Arsenault, B.

1998-01-01

The purpose of this analysis is to validate the Reactor Fuelling Simulation Program (RFSP) using the simple-cell model (SCM) history-based method in a startup simulation following a reactor shutdown period. This study is part of the validation work for history-based calculations, using the WIMS-AECL code with the ENDF/B-V library, and the SCM linked to the RFSP code. In this work, the RFSP code with the SCM history-based method was used to track a 1-year period of the Point Lepreau reactor operating history, that included a 12-day reactor shutdown and subsequent startup. Measured boron and gadolinium concentrations were used in the RFSP simulations, and the predicted values of core reactivity were compared to the reference (pre-shutdown) value. The discrepancies in core reactivity are shown to be better than ±2 milli-k at any time, and better than about ±0.5 milli-k towards the end of the startup transient. The results of this analysis also show that the calculated maximum channel and bundle powers are within an acceptable range during both the core-follow and the reactor startup simulations. (author)

Circuit realization, chaos synchronization and estimation of parameters of a hyperchaotic system with unknown parameters

Directory of Open Access Journals (Sweden)

A. Elsonbaty

2014-10-01

Full Text Available In this article, the adaptive chaos synchronization technique is implemented by an electronic circuit and applied to the hyperchaotic system proposed by Chen et al. We consider the more realistic and practical case where all the parameters of the master system are unknowns. We propose and implement an electronic circuit that performs the estimation of the unknown parameters and the updating of the parameters of the slave system automatically, and hence it achieves the synchronization. To the best of our knowledge, this is the first attempt to implement a circuit that estimates the values of the unknown parameters of chaotic system and achieves synchronization. The proposed circuit has a variety of suitable real applications related to chaos encryption and cryptography. The outputs of the implemented circuits and numerical simulation results are shown to view the performance of the synchronized system and the proposed circuit.

Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

International Nuclear Information System (INIS)

Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

2016-01-01

USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

78 FR 60260 - Order of the Commodity Futures Trading Commission Relating to the Continuation, Shutdown, and...

Science.gov (United States)

2013-10-01

..., cyber security incidents or financial emergencies throughout a lapse in appropriations. C. Extension of...) price discovery; (4) sound risk management practices; and (5) other public interest considerations. The... malfunctions, cyber-security incidents, and financial emergencies shall continue during a shutdown. The...

78 FR 79709 - Duke Energy Florida, Inc., Crystal River Unit 3 Nuclear Generating Plant Post-Shutdown...

Science.gov (United States)

2013-12-31

... NUCLEAR REGULATORY COMMISSION [Docket No. 50-302; NRC-2013-0283] Duke Energy Florida, Inc., Crystal River Unit 3 Nuclear Generating Plant Post-Shutdown Decommissioning Activities Report AGENCY: Nuclear Regulatory Commission (NRC). ACTION: Notice of receipt; availability; public meeting; and request...

Upgrade of the Inner Tracking System of ALICE

CERN Document Server

Kofarago, Monika

2015-01-01

The upgrade of the Inner Tracking System (ITS) of ALICE is planned for the second long shutdown of the LHC in 2019-2020. The ALICE physics program after the shutdown requires the ITS to have improved tracking capabilities and improved impact parameter resolution at very low transverse momentum, as well as a substantial increase in the readout rate. To fulfill these requirements the current ITS will be replaced by seven layers of Monolithic Active Pixel Sensors. The new detector will be moved as close as 23 mm to the interaction point and will have a significantly reduced material budget. Several prototypes of the sensor have been developed to test different aspects of the sensor design including prototypes with analog and digital readout, as well as small and final-size sensors. These prototypes have been thoroughly characterized both in laboratory tests and at test beam facilities including studies on the radiation hardness of the sensors. This contribution gives an overview of the current status of the rese...

Elementary calculation of the shutdown delay of a pile; Calcul elementaire de la periode d'extinction d'une pile

Energy Technology Data Exchange (ETDEWEB)

Yvon, J

1949-04-01

This study analyzes theoretically the progress of the shutdown of a nuclear pile (reactor) when a cadmium rod is introduced instantaneously. For simplification reasons, the environment of the pile is considered as homogenous and only thermal neutrons are considered (delayed neutrons are neglected). Calculation is made first for a plane configuration (plane vessel, plane multiplier without reflector, and plane multiplier with reflector), and then for a cylindrical configuration (multiplier without reflector, multiplier with infinitely thick reflector, finite cylindrical piles without reflector and with reflector). The self-sustain conditions are calculated for each case and the multiplication length and the shutdown delay are deduced. (J.S.)

Identification of human-induced initiating events in the low power and shutdown operation using the commission error search and assessment method

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Chan; Kim, Jong Hyun [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of)

2015-03-15

Human-induced initiating events, also called Category B actions in human reliability analysis, are operator actions that may lead directly to initiating events. Most conventional probabilistic safety analyses typically assume that the frequency of initiating events also includes the probability of human-induced initiating events. However, some regulatory documents require Category B actions to be specifically analyzed and quantified in probabilistic safety analysis. An explicit modeling of Category B actions could also potentially lead to important insights into human performance in terms of safety. However, there is no standard procedure to identify Category B actions. This paper describes a systematic procedure to identify Category B actions for low power and shutdown conditions. The procedure includes several steps to determine operator actions that may lead to initiating events in the low power and shutdown stages. These steps are the selection of initiating events, the selection of systems or components, the screening of unlikely operating actions, and the quantification of initiating events. The procedure also provides the detailed instruction for each step, such as operator's action, information required, screening rules, and the outputs. Finally, the applicability of the suggested approach is also investigated by application to a plant example.

Current status of experimental breeder reactor-II [EBR-II] shutdown planning

International Nuclear Information System (INIS)

McDermott, M. D.; Griffin, C. D.; Michelbacher, J. A.; Earle, O. K.

2000-01-01

The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad

Kinetic parameters for source driven systems

International Nuclear Information System (INIS)

Dulla, S.; Ravetto, P.; Carta, M.; D'Angelo, A.

2006-01-01

The definition of the characteristic kinetic parameters of a subcritical source-driven system constitutes an interesting problem in reactor physics with important consequences for practical applications. Consistent and physically meaningful values of the parameters allow to obtain accurate results from kinetic simulation tools and to correctly interpret kinetic experiments. For subcritical systems a preliminary problem arises for the adoption of a suitable weighting function to be used in the projection procedure to derive a point model. The present work illustrates a consistent factorization-projection procedure which leads to the definition of the kinetic parameters in a straightforward manner. The reactivity term is introduced coherently with the generalized perturbation theory applied to the source multiplication factor ks, which is thus given a physical role in the kinetic model. The effective prompt lifetime is introduced on the assumption that a neutron generation can be initiated by both the fission process and the source emission. Results are presented for simplified configurations to fully comprehend the physical features and for a more complicated highly decoupled system treated in transport theory. (authors)

CAREM-25: Residual heat removal system

International Nuclear Information System (INIS)

Arvia, Roberto P.; Coppari, Norberto R.; Gomez de Soler, Susana M.; Ramilo, Lucia B.

2000-01-01

The objective of this work was the definition and consolidation of the residual heat removal system for the CAREM 25 reactor. The function of this system is cool down the primary circuit, removing the core decay heat from hot stand-by to cold shutdown and during refueling. In addition, this system heats the primary water from the cold shutdown condition to hot stand-by condition during the reactor start up previous to criticality. The system has been designed according to the requirements of the standards: ANSI/ANS 51.1 'Nuclear safety criteria for the design of stationary PWR plants'; ANSI/ANS 58.11 'Design criteria for safe shutdown following selected design basis events in light water reactors' and ANSI/ANS 58.9 'Single failure criteria for light water reactor safety-related fluid systems'. The suggested design fulfills the required functions and design criteria standards. (author)

Nuclear reactor shutdown control rod assembly

International Nuclear Information System (INIS)

Bilibin, K.

1988-01-01

This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

Dynamic Parameter Identification of Hydrodynamic Bearing-Rotor System

Directory of Open Access Journals (Sweden)

Zhiqiang Song

2015-01-01

Full Text Available A new method called modal parameter genetic time domain identification was employed to study the characteristics of the bearing-rotor system. A multifrequency signal decomposition technology to identify the main components of the measured signal and reject the image mode produced by noise has been used. The first- and second-order natural frequency and damping ratios of the shaft system are identified. Furthermore, because of the deficiency of the traditional least square method, a new genetic identification method to identify the bearing dynamic characteristic parameters has been proposed. The method has been effective albeit with few testing points and operation cases. The derivation of oil-film dynamic coefficients could also provide a basis for shaft system natural vibration characteristic and vibration response analysis. Using the identified dynamic coefficients as the supporting condition, the shaft system modal characteristics were studied. The calculated first- and second-order natural frequencies match quite well those obtained from the modal parameter identification. It was proved that the modal parameter and physical parameter identification methods utilized in this paper are reasonable.

Concept design of multipurpose gamma irradiator ISG-500 instrumentation and control system

International Nuclear Information System (INIS)

Dian F Atmoko; Sutomo B; Ikhsan S; A Suntoro

2010-01-01

Has been concept designed of multipurpose 2 x 250 kCi gamma irradiator instrumentation and control system (ICS). The problem in ICS of irradiator is How to get similar of dose rate and start-up/shut down mechanism with highest safety factor. The concept designed of ICS had of tree parameter such as safety, operation and security. The tree of parameter used to start-up and shut-down in irradiator installation with interlock system connection to guarantee of safety. Similar of dose rate obtained by controlled of exposure time witch stopped of carrier conveyor in point of stopped carrier and for delay time, with speed of moved motor carrier to set in constant speed. (author)

Reserves for shutdown/dismantling and disposal in nuclear technology. Theses and recommendations on reform options; Rueckstellungen fuer Stilllegung/Rueckbau und Entsorgung im Atombereich. Thesen und Empfehlungen zu Reformoptionen

Energy Technology Data Exchange (ETDEWEB)

Meyer, Bettina [Forum Oekologisch-Soziale Marktwirtschaft e.V. (FOeS), Berlin (Germany). Green Budget Germany (GBG)

2012-04-11

The study on reserves for shutdown, dismantling and disposal of nuclear facilities covers the following topics: cost for shutdown, dismantling and disposal and amount and transparency of nuclear reserves, solution by y stock regulated by public law for long-term liabilities, and improvement of the protection in the event of insolvency for the remaining EVU reserves for short- and intermediate-term liabilities. The appendix includes estimations and empirical values for the cost of shutdown and dismantling, estimation of disposal costs, and a summary of Swiss studies on dismantling and disposal and transfer to Germany.

77 FR 72294 - Reconsideration of Certain New Source and Startup/Shutdown Issues: National Emission Standards...

Science.gov (United States)

2012-12-05

... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Parts 60 and 63 [EPA-HQ-OAR-2009-0234; EPA-HQ-OAR-2011-0044; FRL-9733-2] RIN 2060-AR62 Reconsideration of Certain New Source and Startup/Shutdown Issues: National Emission Standards for Hazardous Air Pollutants From Coal- and Oil-Fired Electric Utility Steam Generating...

High level waste facilities - Continuing operation or orderly shutdown

International Nuclear Information System (INIS)

Decker, L.A.

1998-04-01

Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed

Downstream passage and impact of turbine shutdowns on survival of silver American Eels at five hydroelectric dams on the Shenandoah River

Science.gov (United States)

Eyler, Sheila; Welsh, Stuart A.; Smith, David R.; Rockey, Mary

2016-01-01

Hydroelectric dams impact the downstream migrations of silver American Eels Anguilla rostrata via migratory delays and turbine mortality. A radiotelemetry study of American Eels was conducted to determine the impacts of five run-of-the-river hydroelectric dams located over a 195-km stretch of the Shenandoah River, Virginia–West Virginia, during fall 2007–summer 2010. Overall, 96 radio-tagged individuals (mean TL = 85.4 cm) migrated downstream past at least one dam during the study. Most American Eels passed dams relatively quickly; over half (57.9%) of the dam passage events occurred within 1 h of reaching a dam, and most (81.3%) occurred within 24 h of reaching the dam. Two-thirds of the dam passage events occurred via spill, and the remaining passage events were through turbines. Migratory delays at dams were shorter and American Eels were more likely to pass via spill over the dam during periods of high river discharge than during low river discharge. The extent of delay in migration did not differ between the passage routes (spill versus turbine). Twenty-eight American Eels suffered turbine-related mortality, which occurred at all five dams. Mortality rates for eels passing through turbines ranged from 15.8% to 40.7% at individual dams. Overall project-specific mortality rates (with all passage routes combined) ranged from 3.0% to 14.3%. To protect downstream-migrating American Eels, nighttime turbine shutdowns (1800–0600 hours) were implemented during September 15–December 15. Fifty percent of all downstream passage events in the study occurred during the turbine shutdown period. Implementation of the seasonal turbine shutdown period reduced cumulative mortality from 63.3% to 37.3% for American Eels passing all five dams. Modifying the turbine shutdown period to encompass more dates in the spring and linking the shutdowns to environmental conditions could provide greater protection to downstream-migrating American Eels.

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

Management of refuelling, modifications and accidental shut-down of nuclear power plant

International Nuclear Information System (INIS)

1996-01-01

This document is the appendix of HAF 0300 (91) 'Code on the Safety of Nuclear Power Plant Operation', which was promulgated by the National Nuclear Safety Administration (NNSA) on March 2, 1994, and has the same legal effect. This appendix is applicable to establish the administrative management procedures for refuelling, modifications and accidental shut-down in the period of operation of pressurized water thermal neutron reactor of nuclear power plants. The NNSA shall be responsible for interpretation of this document

Spent fuel acceptance scenarios devoted to shutdown reactors: A preliminary analysis

International Nuclear Information System (INIS)

Wood, T.W.; Plummer, A.M.; Dippold, D.G.; Short, S.M.

1989-10-01

Spent fuel acceptance schedules and the allocation of federal acceptance capacity among commercial nuclear power reactors have important operational and cost consequences for reactor operators. Alternative allocation schemes were investigated to some extent in DOE's MRS Systems Study. The current study supplements these analyses for a class of acceptance schemes in which the acceptance capacity of the federal radioactive waste management system is allocated principally to shutdown commercial power reactors, and extends the scope of analysis to include considerations of at-reactor cask loading rates. The operational consequences of these schemes for power reactors, as measured in terms of quantity of spent fuel storage requirement above storage pool capacities and number of years of pool operations after last discharge, are estimated, as are the associated utility costs. This study does not attempt to examine the inter-utility equity considerations involved in departures from the current oldest-fuel-first (OFF) allocation rule as specified in the ''Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste.'' In the sense that the alternative allocations are more economically efficient than OFF, however, they approximate the allocations that could result from free exchange of acceptance rights among utilities. Such a process would result in the preservation of inter-utility equity. 13 refs., 9 figs., 9 tabs

Recent experience about the influence of primary coolant and shutdown chemistry on cobalt activity at Beznau NPP

International Nuclear Information System (INIS)

Mailand, I.; Venz, H.

2007-01-01

The Beznau nuclear power plant comprises two identical 380 MWe PWR units, commissioned in 1969 and 1971. The surfaces of the new steam generator tube material, Inconel 690, are the main source of 58 Co. The 60 Co originates predominantly from the Cobalt alloy, Stellite, which is installed in valves and pump bearings because of the very good hardness of this material. By means of optimised shutdown chemistry it is possible to reduce the amount of NiO on the fuel rods, leading to reduced Co-58 peaks in subsequent cycles. The optimised shutdown chemistry during the past few years and especially the strict separation of acid-reducing phase from the acid-oxidising phase as well as the results of studies and the resulting operational experiences are important basics for the actual operation mode of the Beznau NPP. (orig.)

Radiologic states of the WWR-S Bucharest Reactor following definitive shutdown

International Nuclear Information System (INIS)

Garlea, C.; Kelerman, C.; Mocioiu, D.; Garlea, I.

2001-01-01

The definitive shutdown of a reactor raises problems related to the management of the radioactive inventory. To define the radioactive inventory contained in the burned nuclear fuel and in the neutron activated structural materials computation methods are to be used. Besides the radioactive inventory contained in the main block of the reactor, the one due to the primary circuit contaminated mainly with fission products and corrosion products activated in the reactor core, transported and deposed on the components of the cooling primary circuit should be added. Also another component of the radioactive inventory intervenes, namely, the one due to the contamination of the technological rooms used for various operations the nuclear activities (hot cells, pump room, reactor hall, passage ways to the hot cells and for radioactive source, radioisotope and radioactive waste transport). The activities which made used of the neutron and gamma fluxes for radioisotope production, materials irradiation, research, component testing, resulted in radioactive waste, technological or accidental contaminations of the technological rooms of the reactor. Inspections and current repair interventions resulted also in radioactive waste an contaminations. Consequently systematic measurements with qualified equipment dedicated to alpha, beta, gamma contamination measurements as well as to dose rates determinations for the personnel exposed are necessary. Irrespective of the duration of the reactor conservation or shutdown, the radiologic monitoring should continue. This work presents the results obtained by the research group 'Restoration of Nuclear Sites', working with the IFIN-HH, regarding both the radioactive inventory calculation and measurements of contamination of technological rooms and environment in the reactor vicinity

Accelerated maximum likelihood parameter estimation for stochastic biochemical systems

Directory of Open Access Journals (Sweden)

Daigle Bernie J

2012-05-01

Full Text Available Abstract Background A prerequisite for the mechanistic simulation of a biochemical system is detailed knowledge of its kinetic parameters. Despite recent experimental advances, the estimation of unknown parameter values from observed data is still a bottleneck for obtaining accurate simulation results. Many methods exist for parameter estimation in deterministic biochemical systems; methods for discrete stochastic systems are less well developed. Given the probabilistic nature of stochastic biochemical models, a natural approach is to choose parameter values that maximize the probability of the observed data with respect to the unknown parameters, a.k.a. the maximum likelihood parameter estimates (MLEs. MLE computation for all but the simplest models requires the simulation of many system trajectories that are consistent with experimental data. For models with unknown parameters, this presents a computational challenge, as the generation of consistent trajectories can be an extremely rare occurrence. Results We have developed Monte Carlo Expectation-Maximization with Modified Cross-Entropy Method (MCEM2: an accelerated method for calculating MLEs that combines advances in rare event simulation with a computationally efficient version of the Monte Carlo expectation-maximization (MCEM algorithm. Our method requires no prior knowledge regarding parameter values, and it automatically provides a multivariate parameter uncertainty estimate. We applied the method to five stochastic systems of increasing complexity, progressing from an analytically tractable pure-birth model to a computationally demanding model of yeast-polarization. Our results demonstrate that MCEM2 substantially accelerates MLE computation on all tested models when compared to a stand-alone version of MCEM. Additionally, we show how our method identifies parameter values for certain classes of models more accurately than two recently proposed computationally efficient methods

Site Characterization Report ORGDP Diffusion Facilities Permanent Shutdown K-700 Power House and K-27 Switch Yard/Switch House

Energy Technology Data Exchange (ETDEWEB)

Thomas R.J., Blanchard R.D.

1988-06-13

The K-700 Power House area, initially built to supply power to the K-25 gaseous diffusion plant was shutdown and disassembled in the 1960s. This shutdown was initiated by TVA supplying economical power to the diffusion plant complex. As a result of world wide over production of enriched, reactor grade U{sup 235}, the K-27 switch yard and switch house area was placed in standby in 1985. Subsequently, as the future production requirements decreased, the cost of production increased and the separation technologies for other processes improved, the facility was permanently shutdown in December, 1987. This Site Characterization Report is a part of the FY-88 engineering Feasibility Study for placing ORGDP Gaseous Diffusion Process facilities in 'Permanent Shutdown'. It is sponsored by the Department of Energy through Virgil Lowery of Headquarters--Enrichment and through Don Cox of ORO--Enrichment Operations. The primary purpose of these building or site characterization reports is to document, quantify, and map the following potential problems: Asbestos; PCB containing fluids; Oils, coolants, and chemicals; and External contamination. With the documented quantification of the concerns (problems) the Engineering Feasibility Study will then proceed with examining the potential solutions. For this study, permanent shutdown is defined as the securing and/or conditioning of each facility to provide 20 years of safe service with minimal expenditures and, where feasible, also serving DOE's needs for long-term warehousing or other such low-risk use. The K-700 power house series of buildings were either masonry construction or a mix of masonry and wood. The power generating equipment was removed and sold as salvage in the mid 1960s but the buildings and auxiliary equipment were left intact. The nine ancillary buildings in the power house area use early in the Manhattan Project for special research projects, were left intact minus the original special equipment

Electrospun Nanofibers for Sandwiched Polyimide/Poly (vinylidene fluoride)/Polyimide Separators with the Thermal Shutdown Function

International Nuclear Information System (INIS)

Wu, Dezhi; Shi, Chuan; Huang, Shaohua; Qiu, Xiaochun; Wang, Huan; Zhan, Zhan; Zhang, Peng; Zhao, Jinbao; Sun, Daoheng; Lin, Liwei

2015-01-01

Nanofibers fabricated by the electrospinning process have been used to construct sandwich-type Polyimide/Poly (vinylidene fluoride)/Polyimide (PI/PVDF/PI) separators with the thermal shutdown function for lithium ion batteries. This architecture uses the good thermal stability of PI as the top and bottom structure layers. Under high temperature operations, the middle layer made of PVDF nanofibers can melt and form a pore-free film to shut down the battery operation. The electrolyte uptake and ionic conductivity of the PI/PVDF/PI separator are superior to those of commercial polyolefin separators at 476% and 3.46 mS cm −1 , respectively, resulting better battery performances in terms of impedance, discharge capacity and cycle life. Under high temperature treatments above 170 °C, the self-shutdown function of the PI/PVDF/PI has been observed within 10 minutes, which could serve as the safety mechanism to defend the thermal runaway issue of lithium ion batteries. The effects of heating temperature and different time on the morphologies of each layer and electrolyte uptake of the separator are characterized as well

Development of a safety parameter supervision system for Angra-1

International Nuclear Information System (INIS)

Silva, R.A. da; Thome Filho, Z.D.; Schirru, R.; Martinez, A.S.; Oliveira, L.F.S. de

1986-01-01

The Safety Parameter Supervision System (SSPS) which is a computerized system for monitoring essential parameters in real time, determining the safety status and emergency procedures for returning normal reactor operation, in case of an anomaly occurrence, is presented. The SSPS consists of three sub-systems: Integrated parameter monitoring system which gives to operators an integrated vision of values of a parameter set, able to detect any deviation of normal reactor operation; safety critical function system which evaluates safety status in terms of a safety critical function set appointed in advance, and in case of violation of any critical function, it initiates the adequate emergency procedure to return normal operation; and safety parameter computer system which carries out the arquirement of analogic and digital control signals of nuclear power plant. (M.C.K.) [pt

Improving the security of optoelectronic delayed feedback system by parameter modulation and system coupling

Science.gov (United States)

Liu, Lingfeng; Miao, Suoxia; Cheng, Mengfan; Gao, Xiaojing

2016-02-01

A coupled system with varying parameters is proposed to improve the security of optoelectronic delayed feedback system. This system is coupled by two parameter-varied optoelectronic delayed feedback systems with chaotic modulation. Dynamics performance results show that this system has a higher complexity compared to the original one. Furthermore, this system can conceal the time delay effectively against the autocorrelation function and delayed mutual information method and can increase the dimension space of secure parameters to resist brute-force attack by introducing the digital chaotic systems.

PC based 8-parameter data acquisition system

International Nuclear Information System (INIS)

Gupta, J.D.; Naik, K.V.; Jain, S.K.; Pathak, R.V.; Suman, B.

1989-01-01

Multiparameter data acquisition (MPA) systems which analyse nuclear events with respect to more than one property of the event are essential tools for the study of some complex nuclear phenomena requiring analysis of time coincident spectra. For better throughput and accuracy each parameter is digitized by its own ADC. A stand alone low cost IBM PC based 8-parameter data acquisition system developed by the authors makes use of Address Recording technique for acquiring data from eight 12 bit ADC's in the PC Memory. Two memory buffers in the PC memory are used in ping-pong fashion so that data acquisition in one bank and dumping of data onto PC disk from the other bank can proceed simultaneously. Data is acquired in the PC memory through DMA mode for realising high throughput and hardware interrupt is used for switching banks for data acquisition. A comprehensive software package developed in Turbo-Pascal offers a set of menu-driven interactive commands to the user for setting-up system parameters and control of the system. The system is to be used with pelletron accelerator. (author). 5 figs

Development of Abnormal Operating Strategies for Loss of Ultimate Heat Sink (LOUHS) at Shutdown Mode in Westinghouse Type Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang Jun [KHNP CRI, Daejeon (Korea, Republic of); Hwang, Su-Hyun [FNC Tech. Co., Yongin (Korea, Republic of)

2016-10-15

Loss of all AC power is classified as one of multiple failure accident by regulatory guide of Korean accident management program. Therefore we need develop strategies for the abnormal operating procedure both of power operating and shutdown mode. This paper developed abnormal operating guideline for loss of all AC power by analysis of accident scenario in pressurized water reactor. This paper analyzed the extended SBO in shutdown operating mode and developed the operating strategy of the abnormal operation procedure. Operator action for the emergency are not required to take in 500 minutes and 60 minutes in intact and opened RCS state respectively.

Estimating model parameters in nonautonomous chaotic systems using synchronization

International Nuclear Information System (INIS)

Yang, Xiaoli; Xu, Wei; Sun, Zhongkui

2007-01-01

In this Letter, a technique is addressed for estimating unknown model parameters of multivariate, in particular, nonautonomous chaotic systems from time series of state variables. This technique uses an adaptive strategy for tracking unknown parameters in addition to a linear feedback coupling for synchronizing systems, and then some general conditions, by means of the periodic version of the LaSalle invariance principle for differential equations, are analytically derived to ensure precise evaluation of unknown parameters and identical synchronization between the concerned experimental system and its corresponding receiver one. Exemplifies are presented by employing a parametrically excited 4D new oscillator and an additionally excited Ueda oscillator. The results of computer simulations reveal that the technique not only can quickly track the desired parameter values but also can rapidly respond to changes in operating parameters. In addition, the technique can be favorably robust against the effect of noise when the experimental system is corrupted by bounded disturbance and the normalized absolute error of parameter estimation grows almost linearly with the cutoff value of noise strength in simulation

On the principles of the determination of the safe shut-down earthquake for nuclear power plants in Austria

International Nuclear Information System (INIS)

Drimmel, J.

1976-01-01

At present no legal guide lines exist in Austria for the determination of the Safe Shut-Down Earthquake. According to experience, the present requirements for a nuclear power plant site are the following: It must be free of marked tectonic faults and it must never have been situated within the epicentral region of a strong earthquake. The maximum expected earthquake and the Safe Shut-Down Earthquake respectively, are fixed by the aid of a frequency map of strong earthquakes and a map of extreme earthquake intensities in Austria based on macroseismic data since 1201 A.D. The corresponding values of acceleration will be prescribed according to the state of science, but must at least be 0.10 g for the horizontal and 0.05 g for the vertical component of acceleration at the basement

Reference clock parameters for digital communications systems applications

Science.gov (United States)

Kartaschoff, P.

1981-01-01

The basic parameters relevant to the design of network timing systems describe the random and systematic time departures of the system elements, i.e., master (or reference) clocks, transmission links, and other clocks controlled over the links. The quantitative relations between these parameters were established and illustrated by means of numerical examples based on available measured data. The examples were limited to a simple PLL control system but the analysis can eventually be applied to more sophisticated systems at the cost of increased computational effort.

Prism reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-08-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

Simulation of an SBLOCA Test of Shutdown Cooling System Line Break with the SMARTITL Facility using the MARS-KS Code

Energy Technology Data Exchange (ETDEWEB)

Cho, Yeon-Sik; Suh, Jae-Seung [System Engineering and Technology, Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sung-Uk; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

An LBLOCA (Large-Break Loss of Coolant Accident) was inherently eliminated in the design stage. The SMART design has a thermal power of 330MW. Its core exit temperature and pressurizer pressure are 323 .deg. C and 15MPa during normal operating conditions, respectively. An integral-effect test loop for SMART (SMARTITL), called FESTA (Facility for Experimental Simulation of Transients and Accidents), was designed to simulate the integral thermal-hydraulic behavior of SMART. The objectives of SMART-ITL are to investigate and understand the integral performance of reactor systems and components, and the thermal-hydraulic phenomena occurring in the system during normal, abnormal, and emergency conditions, and to verify the system safety during various design basis events of SMART. SMART-ITL with four steam generators and PRHRS, has an advantage for a multi-loop effect compared with VISTA-ITL with a single loop. The integral-effect test data will also be used to validate the related thermal-hydraulic models of the safety analysis code such as TASS/SMR-S which is used for a performance and accident analysis of the SMART design. In addition, a scoping analysis on the scaling difference between the standard design of SMART and the basic design of SMART-ITL was performed for an SBLOCA (Small-Break Loss of Coolant Accident) scenario using a best-estimate safety analysis code, MARS-KS. This paper introduces a comparison of an SBLOCA test of a shutdown cooling system line break using SMART-ITL with its post-test calculation using the MARS-KS code. An SBLOCA test and its post-test calculation were successfully performed using the SMART-ITL facility and MARS-KS code. The SBLOCA break is a guillotine break, and its location is on the SCS line (nozzle part of the RCP suction). The steady-state conditions were achieved to satisfy the initial test conditions presented in the test requirement and its boundary conditions were properly simulated.

Network Flow Simulation of Fluid Transients in Rocket Propulsion Systems

Science.gov (United States)

Bandyopadhyay, Alak; Hamill, Brian; Ramachandran, Narayanan; Majumdar, Alok

2011-01-01

Fluid transients, also known as water hammer, can have a significant impact on the design and operation of both spacecraft and launch vehicle propulsion systems. These transients often occur at system activation and shutdown. The pressure rise due to sudden opening and closing of valves of propulsion feed lines can cause serious damage during activation and shutdown of propulsion systems. During activation (valve opening) and shutdown (valve closing), pressure surges must be predicted accurately to ensure structural integrity of the propulsion system fluid network. In the current work, a network flow simulation software (Generalized Fluid System Simulation Program) based on Finite Volume Method has been used to predict the pressure surges in the feed line due to both valve closing and valve opening using two separate geometrical configurations. The valve opening pressure surge results are compared with experimental data available in the literature and the numerical results compared very well within reasonable accuracy (< 5%) for a wide range of inlet-to-initial pressure ratios. A Fast Fourier Transform is preformed on the pressure oscillations to predict the various modal frequencies of the pressure wave. The shutdown problem, i.e. valve closing problem, the simulation results are compared with the results of Method of Characteristics. Most rocket engines experience a longitudinal acceleration, known as "pogo" during the later stage of engine burn. In the shutdown example problem, an accumulator has been used in the feed system to demonstrate the "pogo" mitigation effects in the feed system of propellant. The simulation results using GFSSP compared very well with the results of Method of Characteristics.

Evaluation of the reliability of the protection system of 1300 MWE PWR'S

International Nuclear Information System (INIS)

Blin, A.

1990-01-01

An assesment of the reliability of the Digital Integrated Protection System (SPIN) of the 1300 MWe type french reactors has been carried out by treating an example: the emergency shutdown, which can be called upon by several initiating events. The whole chain, from sensors to breakers and control rods, is taken into account. The reliability parameters used for the quantification are evaluated essentially from the experience feedback of french reactors. The not wellknown parameters being the common cause failure rates of electronic components and the efficiency rate of the self-tests, the results of the study are then presented in a parametric form, according to these two factors

Site investigations, design, construction, operation, shutdown and surveillance of repositories for low- and intermediate-level radioactive wastes in rock cavities

International Nuclear Information System (INIS)

1984-01-01

The report provides an overview and technical guidelines for considerations and for activities to be undertaken for safety assessment, site investigations, design, construction, operation, shutdown and surveillance of repositories for the disposal of low- and intermediate-level radioactive wastes in rock cavities. A generalized sequence of investigations is introduced which proceeds through region and site selection to the stage where the site is confirmed by detailed geoscientific investigations as being suitable for a waste repository. The different procedures and somewhat specific investigative needs with respect to existing mines are dealt with separately. General design, as well as specific requirements with respect to the different stages of design and construction, are dealt with. A review of activities related to the operational and post-operational stages of repositories in rock cavities is presented. The report describes in general terms the procedures related to different stages of disposal operation; also the conditions for shutdown together with essential shutdown and sealing activities and the related safety assessment requirements. Guidance is also given on the surveillance programme which will allow for inspection, testing, maintenance and security of a disposal facility during the operational phase, as well as for the post-operational stage for periods determined as necessary by the national authorities

Sensitivity analysis in multi-parameter probabilistic systems

International Nuclear Information System (INIS)

Walker, J.R.

1987-01-01

Probabilistic methods involving the use of multi-parameter Monte Carlo analysis can be applied to a wide range of engineering systems. The output from the Monte Carlo analysis is a probabilistic estimate of the system consequence, which can vary spatially and temporally. Sensitivity analysis aims to examine how the output consequence is influenced by the input parameter values. Sensitivity analysis provides the necessary information so that the engineering properties of the system can be optimized. This report details a package of sensitivity analysis techniques that together form an integrated methodology for the sensitivity analysis of probabilistic systems. The techniques have known confidence limits and can be applied to a wide range of engineering problems. The sensitivity analysis methodology is illustrated by performing the sensitivity analysis of the MCROC rock microcracking model

Management-retrieval code system of fission barrier parameter sub-library

International Nuclear Information System (INIS)

Zhang Limin; Su Zongdi; Ge Zhigang

1995-01-01

The fission barrier parameter (FBP) library, which is a sub-library of Chinese Evaluated Nuclear Parameter library (CENPL), stores various popular used fission barrier parameters from different historical period, and could retrieve the required fission barrier parameters by using the management retrieval code system of the FBP sub-library. The function, feature and operation instruction of the code system are described briefly

PSA-operations synergism for the advanced test reactor shutdown operations PSA

International Nuclear Information System (INIS)

Atkinson, S.A.

1996-01-01

The Advanced Test Reactor (ATR) Probabilistic Safety Assessment (PSA) for shutdown operations, cask handling, and canal draining is a successful example of the importance of good PSA-operations synergism for achieving a realistic and accepted assessment of the risks and for achieving desired risk reduction and safety improvement in a best and cost-effective manner. The implementation of the agreed-upon upgrades and improvements resulted in the reductions of the estimated mean frequency for core or canal irradiated fuel uncovery events, a total reduction in risk by a factor of nearly 1000 to a very low and acceptable risk level for potentially severe events

Basis for Interim Operation for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2003-01-01

This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

Control interlock and monitoring system for 80 KW IOT based RF power amplifier system at 505.812 MHz for Indus-2

International Nuclear Information System (INIS)

Kumar, Gautam; Deo, R.K.; Jain, M.K.; Bagre, Sunil; Hannurkar, P.R.

2013-01-01

For 80 kW inductive output tube (IOT) based RF power amplifier system at 505.812 MHz for Indus-2, a control, interlock and monitoring system is realized. This is to facilitate proper start-up and shutdown of the amplifier system, monitor various parameters to detect any malfunction during its operation and to bring the system in a safe stage, thereby assuring reliable operation of the amplifier system. This high power amplifier system incorporates interlocks such as cooling interlocks, various voltage and current interlocks and time critical RF interlocks. Processing of operation sequence, cooling interlocks and various voltage and current interlocks have been realized by using Siemens make S7-CPU-315-2DP (CPU) based programmable logic controller (PLC) system. While time critical or fast interlocks have been realized by using Siemens make FPGA based Boolean Co-processor FM-352-5 which operates in standalone mode. Siemens make operating panel OP277 6'' is being used as a human machine interface (HMI) device for command, data, alarm generation and process parameter monitoring. (author)

Shutdown decay heat removal analysis: Plant case studies and special issues: Summary report

International Nuclear Information System (INIS)

Ericson, D.M. Jr.; Cramond, W.R.; Sanders, G.A.; Hatch, S.W.

1989-04-01

Shutdown Decay Heat Removal Requirements has been designated as Unresolved Safety Issue (USI) A-45. The overall objectives of the USI A-45 program were to evaluate the safety adequacy of decay heat removal (DHR) systems in existing light water reactor nuclear power plants and to assess the value and impact (benefit-cost) of alternative measures for improving the overall reliability of the DHR function. To provide the technical data required to meet these objectives a program was developed that examined the state of DHR system reliability in a sample of existing plants. This program identified potential vulnerabilities and identified and established the feasibility of potential measures to improve the reliability of the DHR function. A value/impact (V/I) analysis of the more promising of such measures was conducted and documented. This report summarizes those studies. In addition, because of the evolving nature of V/I analyses in support of regulation, a number of supporting studies related to appropriate procedures and measures for the V/I analyses were also conducted. These studies are also summarized herein. This report only summarizes findings of technical studies performed by Sandia National Laboratories as part of the program to resolve this issue. 46 refs., 7 figs., 124 tabs

Human Reliability Analysis. Applicability of the HRA-concept in maintenance shutdown

International Nuclear Information System (INIS)

Obenius, Aino

2007-08-01

Probabilistic Safety Analysis (PSA) is performed for Swedish nuclear power plants in order to make predictions and improvements of system safety. The analysis of the Three Mile Island and Chernobyl accidents contributed to broaden the approach to nuclear power plant safety. A system perspective focusing on the interaction between aspects of Man, Technology and Organization (MTO) emerged in addition to the development of Human Factors knowledge. To take the human influence on the technical system into consideration when performing PSAs, a Human Reliability Analysis (HRA) is performed. PSA is performed for different stages and plant operating states, and the current state of Swedish analyses is Low power and Shutdown (LPSD), also called Shutdown PSA (SPSA). The purpose of this master's thesis is to describe methods and basic models used when analysing human reliability for the LPSD state. The following questions are at issue: 1. How can the LPSD state be characterised and defined? 2. What is important to take into consideration when performing a LPSD HRA? 3. How can human behaviour be modelled for a LPSD risk analysis? 4. According to available empirical material, how are the questions above treated in performed analysis of human operation during LPSD? 5. How does the result of the questions above affect the way methods for analysis of LPSD could and/or should be developed? The procedure of this project has mainly consisted of literature studies of available theory for modelling of human behaviour and risk analysis of the LPSD state. This study regards analysis of planned outages when maintenance, fuel change, tests and inspections are performed. The outage period is characterised by planned maintenance activities performed in rotating 3-shifts, around the clock, as well as many of the persons performing work tasks on the plant being external contractors. The working conditions are characterised by stress due to heat, radiation and physically demanding or monotonous

An approach of parameter estimation for non-synchronous systems

International Nuclear Information System (INIS)

Xu Daolin; Lu Fangfang

2005-01-01

Synchronization-based parameter estimation is simple and effective but only available to synchronous systems. To come over this limitation, we propose a technique that the parameters of an unknown physical process (possibly a non-synchronous system) can be identified from a time series via a minimization procedure based on a synchronization control. The feasibility of this approach is illustrated in several chaotic systems

Towards automatic parameter tuning of stream processing systems

KAUST Repository

Bilal, Muhammad; Canini, Marco

2017-01-01

for automating parameter tuning for stream-processing systems. Our framework supports standard black-box optimization algorithms as well as a novel gray-box optimization algorithm. We demonstrate the multiple benefits of automated parameter tuning in optimizing

Slope parameters of ππ-system

International Nuclear Information System (INIS)

Isaev, P.S.; Osipov, A.A.

1984-01-01

The slope parameters of the ππ-system are calculated in the framework of the superconductor-tupe quark model. The analogous calculations are made for πK-system. The amplitudes are obtained by using the box quark diagrams and tree diagrams with the intermediate scalar epsilon(700), Ssup(x)(975), K tilde (1350) mesons and vector rho(770), K* (892) mesons

An optical system for controlling ion source parameters

International Nuclear Information System (INIS)

Zhang Baifang; Liu Zhenhao; Jiang Yi; Xu Zhengjia

1999-01-01

An optical control system used for adjusting the source's parameters of an ion separator is described. There are two slice microcomputers at HV terminal and the ground respectively. These microcomputers communicate each other with the full-duplex mode through two pieces of optical fiber, in which many parameters are time-share transmitted in the form of optical pulse. This system can stabilize the arc current and temperature, adjust and display all parameters and has safe-guard ability. At HV terminal, the optical coupling technique is used for connecting the CPU and the ion source, and at the ground the CPU can communicate with a control microcomputer

An internet-based telemonitoring system of multiphysiological parameters.

Science.gov (United States)

Shuicai, Wu; Haomin, Li; Fangfang, Du; Yanping, Bai; Song, Zhang

2007-08-01

The purpose of this research was to design and realize a real-time tele-monitoring system with multiphysiological parameters using the Internet. Both the Client/Server (C/S) mode and Peer-to-Peer (P2P) mode were used in the system's network communication. The C/S mode is used to upload, retrieve, and download physiological data. The P2P mode provides realtime tele-monitoring and video chatting between doctors and patients. Experiment results show that P2P technology could efficiently improve the transmission speed of the physiological parameters. This study demonstrates an effective method of remote monitoring of physiological parameters in real time.

Identification of System Parameters by the Random Decrement Technique

DEFF Research Database (Denmark)

Brincker, Rune; Kirkegaard, Poul Henning; Rytter, Anders

1991-01-01

-Walker equations and finally, least-square fitting of the theoretical correlation function. The results are compared to the results of fitting an Auto Regressive Moving Average (ARMA) model directly to the system output from a single-degree-of-freedom system loaded by white noise.......The aim of this paper is to investigate and illustrate the possibilities of using correlation functions estimated by the Random Decrement Technique as a basis for parameter identification. A two-stage system identification system is used: first, the correlation functions are estimated by the Random...... Decrement Technique, and then the system parameters are identified from the correlation function estimates. Three different techniques are used in the parameter identification process: a simple non-parametric method, estimation of an Auto Regressive (AR) model by solving an overdetermined set of Yule...

New Fast Beam Conditions Monitoring (BCM1F) system for CMS

Science.gov (United States)

Zagozdzinska, A. A.; Bell, A. J.; Dabrowski, A. E.; Hempel, M.; Henschel, H. M.; Karacheban, O.; Przyborowski, D.; Leonard, J. L.; Penno, M.; Pozniak, K. T.; Miraglia, M.; Lange, W.; Lohmann, W.; Ryjov, V.; Lokhovitskiy, A.; Stickland, D.; Walsh, R.

2016-01-01

The CMS Beam Radiation Instrumentation and Luminosity (BRIL) project is composed of several systems providing the experiment protection from adverse beam conditions while also measuring the online luminosity and beam background. Although the readout bandwidth of the Fast Beam Conditions Monitoring system (BCM1F—one of the faster monitoring systems of the CMS BRIL), was sufficient for the initial LHC conditions, the foreseen enhancement of the beams parameters after the LHC Long Shutdown-1 (LS1) imposed the upgrade of the system. This paper presents the new BCM1F, which is designed to provide real-time fast diagnosis of beam conditions and instantaneous luminosity with readout able to resolve the 25 ns bunch structure.

Robustness of dynamic systems with parameter uncertainties

CERN Document Server

Balemi, S; Truöl, W

1992-01-01

Robust Control is one of the fastest growing and promising areas of research today. In many practical systems there exist uncertainties which have to be considered in the analysis and design of control systems. In the last decade methods were developed for dealing with dynamic systems with unstructured uncertainties such as HOO_ and £I-optimal control. For systems with parameter uncertainties, the seminal paper of V. L. Kharitonov has triggered a large amount of very promising research. An international workshop dealing with all aspects of robust control was successfully organized by S. P. Bhattacharyya and L. H. Keel in San Antonio, Texas, USA in March 1991. We organized the second international workshop in this area in Ascona, Switzer­ land in April 1992. However, this second workshop was restricted to robust control of dynamic systems with parameter uncertainties with the objective to concentrate on some aspects of robust control. This book contains a collection of papers presented at the International W...

DAQ system for low density plasma parameters measurement

International Nuclear Information System (INIS)

Joshi, Rashmi S.; Gupta, Suryakant B.

2015-01-01

In various cases where low density plasmas (number density ranges from 1E4 to 1E6 cm -3 ) exist for example, basic plasma studies or LEO space environment measurement of plasma parameters becomes very critical. Conventional tip (cylindrical) Langmuir probes often result into unstable measurements in such lower density plasma. Due to larger surface area, a spherical Langmuir probe is used to measure such lower plasma densities. Applying a sweep voltage signal to the probe and measuring current values corresponding to these voltages gives V-I characteristics of plasma which can be plotted on a digital storage oscilloscope. This plot is analyzed for calculating various plasma parameters. The aim of this paper is to measure plasma parameters using a spherical Langmuir probe and indigenously developed DAQ system. DAQ system consists of Keithley source-meter and a host system connected by a GPIB interface. An online plasma parameter diagnostic system is developed for measuring plasma properties for non-thermal plasma in vacuum. An algorithm is developed using LabVIEW platform. V-I characteristics of plasma are plotted with respect to different filament current values and different locations of Langmuir probe with reference to plasma source. V-I characteristics is also plotted for forward and reverse voltage sweep generated programmatically from the source meter. (author)

Behavior of antimony isotopes in the primary coolant of WWER-1000-type nuclear reactors in NPP Kozloduy during operation and shutdown

International Nuclear Information System (INIS)

Dobrevski, Ivan D.; Zaharieva, Neli N.; Minkova, Katia F.; Gerchev, Nikolay B.

2009-01-01

This paper focuses on the behavior of the antimony isotopes 122 Sb and 124 Sb in the coolant of the WWER reactors in the nuclear power plant Kozloduy (Bulgaria) during operation and shutdown. It is concluded that the chemical properties of their actual precursor, the isotope 121 Sb, determine the behavior of 122 Sb and 124 Sb during operation, load fluctuations, and shutdown as well as during the reactor coolant purification process. It is supposed that differences between the reactor bulk and the core fuel cladding surface chemistry as well as the presence of sub-cooled nucleate boiling at the fuel cladding may create conditions under which a local oxidizing environment may come into existence. (orig.)

Internal fuel motion as an inherent shutdown mechanism for LMFBR accidents: PINEX-3, PINEX-2, and HUT 5-2A experiments

International Nuclear Information System (INIS)

Ferrell, P.C.; Porten, D.R.; Martin, F.J.

1981-01-01

The PINEX-2 experiment verified the concept of axial internal molten fuel motion within annular fuel, representing an inherent shutdown mechanism for hypothetical transient overpower excursions on the order of 5$/s. The PINEX-3 experiment, simulating a 50 cents/s transient overpower, showed that limitations on the effectiveness of fuel motion may arise from freezing of the fuel and blockage of the internal movement. Analysis of these experiments was performed to assess the physical processes that dominate fuel relocation potential and to apply them to prototypic LMFBR pin conditions. Results indicate that internal fuel motion should be reliable as a shutdown mechanism in LMFBR's for a range of reactivity insertion rates beyond presently available experimental data

Cryogenics system: strategy to achieve nominal performance and reliable operation

CERN Document Server

Bremer, J; Casas, J; Claudet, S; Delikaris, D; Delruelle, N; Ferlin, G; Fluder, C; Perin, A; Perinic, G; Pezzetti, M; Pirotte, O; Tavian, L; Wagner, U

2012-01-01

During the LHC operation in 2010 and 2011, the cryogenic system has achieved an availability level fulfilling the overall requirement. To reach this level, the cryogenic system has profited like many other beam-dependent systems from the reduced beam parameters. Therefore, impacts of some failures occurred during the LHC operation were mitigated by using the overcapacity margin, the existing built-in redundancy in between adjacent sector cryogenic plants and the "cannibalization" of spares on two idle cryogenic plants. These two first years of operation were also crucial to identify the weaknesses of the present cryogenic maintenance plan and new issues like SEUs. After the LS1, nominal beam parameters are expected and the mitigated measures will be less effective or not applicable at all. Consequently, a consolidation plan to improve the MTBF and the MTTR of the LHC cryogenic system is under definition. Concerning shutdown periods, the present cryogenic sectorization imposes some restrictions in the type of ...

Optimization Design of Multi-Parameters in Rail Launcher System

Directory of Open Access Journals (Sweden)

Yujiao Zhang

2014-05-01

Full Text Available Today the energy storage systems are still encumbering, therefore it is useful to think about the optimization of a railgun system in order to achieve the best performance with the lowest energy input. In this paper, an optimal design method considering 5 parameters is proposed to improve the energy conversion efficiency of a simple railgun. In order to avoid costly trials, the field- circuit method is employed to analyze the operations of different structural railguns with different parameters respectively. And the orthogonal test approach is used to guide the simulation for choosing the better parameter combinations, as well reduce the calculation cost. The research shows that the proposed method gives a better result in the energy efficiency of the system. To improve the energy conversion efficiency of electromagnetic rail launchers, the selection of more parameters must be considered in the design stage, such as the width, height and length of rail, the distance between rail pair, and pulse forming inductance. However, the relationship between these parameters and energy conversion efficiency cannot be directly described by one mathematical expression. So optimization methods must be applied to conduct design. In this paper, a rail launcher with five parameters was optimized by using orthogonal test method. According to the arrangement of orthogonal table, the better parameters’ combination can be obtained through less calculation. Under the condition of different parameters’ value, field and circuit simulation analysis were made. The results show that the energy conversion efficiency of the system is increased by 71.9 % after parameters optimization.

Community Design Parameters and the Performance of Residential Cogeneration Systems

Directory of Open Access Journals (Sweden)

Hazem Rashed-Ali

2012-11-01

Full Text Available The integration of cogeneration systems in residential and mixed-use communities has the potential of reducing their energy demand and harmful emissions and can thus play asignificant role in increasing their environmental sustainability. This study investigated the impact of selected planning and architectural design parameters on the environmental and economic performances of centralized cogeneration systems integrated into residential communities in U.S.cold climates. Parameters investigated include: 1 density, 2 use mix, 3 street configuration, 4 housing typology, 5 envelope and building systems’ efficiencies, and 6 passive solar energyutilization. The study integrated several simulation tools into a procedure to assess the impact of each design parameter on the cogeneration system performance. This assessment procedure included: developing a base-line model representing typical design characteristics of U.S. residential communities; assessing the cogeneration system’s performance within this model using three performance indicators: percentage of reduction in primary energy use, percentage of reduction in CO2 emissions; and internal rate of return; assessing the impact of each parameter on the system performance through developing 46 design variations of the base-line model representing potential changes in each parameter and calculating the three indicators for each variation; and finally, using a multi-attribute decision analysis methodology to evaluate the relative impact of each parameter on the cogeneration system performance. The study results show that planning parameters had a higher impact on the cogeneration system performance than architectural ones. Also, a significant correlation was found between design characteristics identified as favorable for the cogeneration system performance and those of sustainable residential communities. These include high densities, high use mix, interconnected street networks, and mixing of

Event sequence quantification for a loss of shutdown cooling accident in the GCFR

International Nuclear Information System (INIS)

Frank, M.; Reilly, J.

1979-10-01

A summary is presented of the core-wide sequence of events of a postulated total loss of forced and natural convection decay heat removal in a shutdown Gas-Cooled Fast Reactor (GCFR). It outlines the analytical methods and results for the progression of the accident sequence. This hypothetical accident proceeds in the distinct phases of cladding melting, assembly wall melting and molten steel relocation into the interassembly spacing, and fuel relocation. It identifies the key phenomena of the event sequence and the concerns and mechanisms of both recriticality and recriticality prevention

Identification of metabolic system parameters using global optimization methods

Directory of Open Access Journals (Sweden)

Gatzke Edward P

2006-01-01

Full Text Available Abstract Background The problem of estimating the parameters of dynamic models of complex biological systems from time series data is becoming increasingly important. Methods and results Particular consideration is given to metabolic systems that are formulated as Generalized Mass Action (GMA models. The estimation problem is posed as a global optimization task, for which novel techniques can be applied to determine the best set of parameter values given the measured responses of the biological system. The challenge is that this task is nonconvex. Nonetheless, deterministic optimization techniques can be used to find a global solution that best reconciles the model parameters and measurements. Specifically, the paper employs branch-and-bound principles to identify the best set of model parameters from observed time course data and illustrates this method with an existing model of the fermentation pathway in Saccharomyces cerevisiae. This is a relatively simple yet representative system with five dependent states and a total of 19 unknown parameters of which the values are to be determined. Conclusion The efficacy of the branch-and-reduce algorithm is illustrated by the S. cerevisiae example. The method described in this paper is likely to be widely applicable in the dynamic modeling of metabolic networks.

Verification and validation of the R2Smesh approach for the calculation of high resolution shutdown dose rate distributions

Czech Academy of Sciences Publication Activity Database

Majerle, Mitja; Leichtle, D.; Fischer, U.; Serikov, A.

2012-01-01

Roč. 87, 5-6 (2012), s. 443-447 ISSN 0920-3796 Institutional support: RVO:61389005 Keywords : MCNP * FISPACT * shutdown dose rate Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 0.842, year: 2012

Selection and verification of safety parameters in safety parameter display system for nuclear power plants

International Nuclear Information System (INIS)

Zhang Yuangfang

1992-02-01

The method and results for safety parameter selection and its verification in safety parameter display system of nuclear power plants are introduced. According to safety analysis, the overall safety is divided into six critical safety functions, and a certain amount of safety parameters which can represent the integrity degree of each function and the causes of change are strictly selected. The verification of safety parameter selection is carried out from the view of applying the plant emergency procedures and in the accident man oeuvres on a full scale nuclear power plant simulator

Hybrid computer optimization of systems with random parameters

Science.gov (United States)

White, R. C., Jr.

1972-01-01

A hybrid computer Monte Carlo technique for the simulation and optimization of systems with random parameters is presented. The method is applied to the simultaneous optimization of the means and variances of two parameters in the radar-homing missile problem treated by McGhee and Levine.

Soft sensing of system parameters in membrane distillation

KAUST Repository

Laleg-Kirati, Taous-Meriem

2017-01-01

Various examples of methods and systems are provided for soft sensing of system parameters in membrane distillation (MD). In one example, a system includes a MD module comprising a feed side and a permeate side separated by a membrane boundary layer

Modeling and Parameter Estimation of a Small Wind Generation System