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Sample records for on-line pwr rhr

  1. On-line PWR RHR pump performance testing following motor and impeller replacement

    International Nuclear Information System (INIS)

    DiMarzo, J.T.

    1996-01-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump's 'B' impeller. The spare was installed into the 'B' train. The motor had never been run in the system before. A pump performance test was developed to verify it's operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the 'B' Train showed performance well in excess of the minimum required. The motor that was originally in the 'B' train was similarly overhauled and equipped with 'A' pump's original impeller, re-installed in the 'A' train, and tested. Analysis of the 'A' train results indicate that the RHR pump's performance was also well in excess of the vendors requirements

  2. On-line PWR RHR pump performance testing following motor and impeller replacement

    Energy Technology Data Exchange (ETDEWEB)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  3. Earthquake resistance of residual heat removed (RHR) pump for pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, K.; Honma, T.; Matsubayashi, H.; Inazuka, H.

    1980-01-01

    The present paper deals with the earthquake resistance of the residual heat removed (RHR) pump of single stage double volute type, which is one of the structurally simplest pumps used for pressurized water reactors (PWR). The results of the study can be summarized as follows: (1) Any trouble which can give effect on the functions of the pump at earthquake does not become a problem so long as each part of the pump is of aseismatically rigid structure. (2) Aseismatic tolerance test in the pump's operating condition has shown that the earthquake resistance of the pump at its location has a tolerance about five times the dynamic design acceleration. (3) The pump is provided with an impeller-casing wear ring at the pressure boundary between the suction side pressure and discharge side pressure. This wear ring acts as an underwater bearing when the pump is in operation, and improves the vibration characteristics, particularly damping ratio, of the pump shaft to a great extent to make the pump more aseismatic. (4) In the evaluation of the underwater bearing characteristics of the wear ring, the evaluation accuracy of the vibration characteristics of the pump shaft can be improved by taking into consideration the pressure loss in the wear ring part from the head of the single stage of the pump due to the rotation of the impeller. (author)

  4. Operating function tests of the PWR type RHR pump for engineering safety system under simulated strong ground excitation

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, Kazuhiro; Homma, Toshiaki; Inazuka, Hisashi; Nakajima, Norifumi.

    1979-08-01

    Results are described of operating function verification tests of a PWR RHR pump during an earthquake. Of the active reactor components, the PWR residual heat removal pump was chosen from view points of aseismic classification, safety function, structural complexity and past aseismic tests. Through survey of the service conditions and structure of this pump, seismic test conditions such as acceleration level, simulated seismic wave form and earthquake duration were decided for seismicity of the operating pump. Then, plans were prepared to evaluate vibration chracteristics of the pump and to estimate its aseismic design margins. Subsequently, test facility and instrumentation system were designed and constructed. Experimental results could thus be acquired on vibration characteristics of the pump and its dynamic behavior during different kinds and levels of simulated earthquake. In conclusion: (1) Stiffeners attached to the auxiliary system piping do improve aseismic performance of the pump. (2) The rotor-shaft-bearing system is secure unless it is subjected to transient disturbunces having high frequency content. (3) The motor and pump casing having resonance frequencies much higher than frequency content of the seismic wave show only small amplifications. (4) The RHR pump possesses an aseismic design margin more than 2.6 times the expected ultimate earthquake on design basis. (author)

  5. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1996-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors, because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW el NPP's with PWR and has been successfully tested in a 350 MW el NPP with a PWR. (orig.)

  6. Operating experience with an on-line vibration control system for PWR main coolant pumps

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Vortriede, A.

    1998-01-01

    The main circulation pumps are key components of nuclear power plants with pressurized water reactors (PWRs), because the availability of the main circulation pumps has a direct influence on the availability and electrical output of the entire plant. The on-line automatic vibration control system ASMAS was developed for early failure detection during the normal operation of the main circulation pumps in order to avoid unexpected outages and to establish the possibility of preventive maintenance of the pumps. This system is permanently and successfully operating in three German 1300 MW e1 NPP's with PWR and has been successfully tested in a 350 MW e1 NPP with a PWR. (orig.)

  7. On-line analysis of ETA and organic acids in secondary systems of PWR plants

    International Nuclear Information System (INIS)

    Kurashina, Masahiko; Uzawa, Hideo; Utagawa, Koya; Takaku, Hiroshi

    2005-01-01

    To reduce the iron concentration in the secondary water of plants with pressurized water reactors (PWRs), ethanolamine (ETA) is used as an alkalizing agent in the secondary cycle. An on-line ion chromatography (IC) monitoring system for monitoring concentrations of ETA and anions of organic acids was developed, its performance was evaluated, and verification tests were conducted at an actual PWR plant. It was demonstrated that the concentration of both ETA and anions of organic acids may be successfully monitored by IC in PWR secondary cycle streams alkalized by ETA. (orig.)

  8. On-line thermal margin estimation of a PWR core using a neural network approach

    International Nuclear Information System (INIS)

    Park, Soon Ok; Kim, Hyun Koon; Lee, Seung Hynk; Chang, Soon Heung

    1992-01-01

    A new approach for on-line thermal margin monitoring of a PWR Core is proposed in this paper, where a neural network model is introduced to predict the DNBR values at the given reactor operating conditions. The neural network is learned by the Back Propagation algorithm with the optimized random training data and is tested to investigate the generalized performance for the steady state operating region as well as for the transient situations where DNB is of the primary concern. The test results show that the high level of accuracy in predicting the DNBR can be achieved by the neural network model compared to the detailed code results. An insight has been gained from this study that the neural network model for estimating DNB performance can be a viable tool for on-line thermal margin monitoring of a nuclear power plant

  9. A study on the computerization of secondary side on-line chemistry monitoring system of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Kyung Lin; Lee, Eun Heui [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-01

    A computer system for on-line chemistry monitoring system located in secondary side of PWR plant is under developing. Keithley 500 A mainframe and AMM1A and AIM3A modules are used for data acquisition and scientific and engineering software package of ASYST is used for developing software program. The contents are as follows: (1) Data acquisition and real-time display. The output signals of monitoring chemical sensors are stored in PC showing real-time data display as true values and graphics. (2) Data management and trending graphs. The data stored in PC are outcoming in various graphic mode for data management such as simple trending graphs screen display, time duration plot and histogram plot. (3) Daily basis data manual input. The chemical analysis data of grab sample are stored in PC by manual input for supplement data. (4) Tabular data report preparation. Summarized daily, weekly, monthly, quarterly and yearly reports are prepared with various mode of graphic display. 6 figs, 9 tabs, 8 refs. (Author).

  10. A study on the computerization of secondary side on-line chemistry monitoring system of PWR

    International Nuclear Information System (INIS)

    Yang, Kyung Lin; Lee, Eun Heui

    1994-12-01

    A computer system for on-line chemistry monitoring system located in secondary side of PWR plant is under developing. Keithley 500 A mainframe and AMM1A and AIM3A modules are used for data acquisition and scientific and engineering software package of ASYST is used for developing software program. The contents are as follows: 1) Data acquisition and real-time display. The output signals of monitoring chemical sensors are stored in PC showing real-time data display as true values and graphics. 2) Data management and trending graphs. The data stored in PC are outcoming in various graphic mode for data management such as simple trending graphs screen display, time duration plot and histogram plot. 3) Daily basis data manual input. The chemical analysis data of grab sample are stored in PC by manual input for supplement data. 4) Tabular data report preparation. Summarized daily, weekly, monthly, quarterly and yearly reports are prepared with various mode of graphic display. 6 figs, 9 tabs, 8 refs. (Author)

  11. Hygrometric measurement for on-line monitoring of PWR vessel head penetrations

    International Nuclear Information System (INIS)

    Germain, J.L.; Loisy, F.; Apolzan, S.

    1994-06-01

    In September 1991, a small leak was found on one of the reactor's upper vessel head penetrations. After inspection, other non-throughwall cracks were localized in the lower part of the vessel head adapter in questions. The same type of crack was later found inside some adapters on other French PWR units. After repairs, the safety authorities granted approval to continue unit operation, with the specific provision that a system for ongoing monitoring of the penetrations be set up. Two types of system were selected to detect leaks through any potential cracks: the first is based on nitrogen-13 detection and the second on steam detection. Both systems call for sampling the air in a confined space above the vessel head. The number and distribution of sampling taps in the circuit, and the balancing of their respective flow rates, are factors in proper monitoring of all vessel head penetrations. Gas-injection holes are also installed in the confined space. These holes are used during the sampling system qualification tests to simulate leaks in various positions and calculate the effective performance of the sampling system. Leaks are simulated using a helium-base gas tracer and measuring tracer concentrations in the sampling system. The system for measuring steam levels in air samples uses chilled-mirror hygrometers. A microcomputer takes regular readings, drives the various automatic functions of the measurement system and automatically analyses the readings so as to monitor operations and trigger an alarm at the first sign of a leak. This system has now been installed for a year and a half on three French PWR units and is functioning satisfactorily. (authors). 5 figs

  12. On-line measurement of gaseous iodine species during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

  13. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2015-01-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  14. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  15. On-line fission products measurements during a PWR severe accident: the French DECA-PF project

    Energy Technology Data Exchange (ETDEWEB)

    Ducros, G.; Allinei, P.G.; Roure, C. [CEA, DEN, F-13108 Saint-Paul-lez-Durance, (France); Rozel, C. [EDF SEPTEN, 12-14 Avenue Dutrievoz, F-69628, Villeurbanne, (France); Blanc De Lanaute, N. [CANBERRA, 1 rue des Herons, F-78182, Saint Quentin en Yvelines, (France); Musoyan, G. [AREVA, Tour AREVA, 1 place Jean Millier, F-92084 Paris La Defense Cedex, (France)

    2015-07-01

    Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, funded in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied

  16. A Study on Structured Simulation Framework for Design and Evaluation of Human-Machine Interface System -Application for On-line Risk Monitoring for PWR Nuclear Power Plant-

    International Nuclear Information System (INIS)

    Zhan, J.; Yang, M.; Li, S.C.; Peng, M.J.; Yan, S.Y.; Zhang, Z.J.

    2006-01-01

    The operators in the main control room of Nuclear Power Plant (NPP) need to monitor plant condition through operation panels and understand the system problems by their experiences and skills. It is a very hard work because even a single fault will cause a large number of plant parameters abnormal and operators are required to perform trouble-shooting actions in a short time interval. It will bring potential risks if operators misunderstand the system problems or make a commission error to manipulate an irrelevant switch with their current operation. This study aims at developing an on-line risk monitoring technique based on Multilevel Flow Models (MFM) for monitoring and predicting potential risks in current plant condition by calculating plant reliability. The proposed technique can be also used for navigating operators by estimating the influence of their operations on plant condition before they take an action that will be necessary in plant operation, and therefore, can reduce human errors. This paper describes the risk monitoring technique and illustrates its application by a Steam Generator Tube Rupture (SGTR) accident in a 2-loop Pressurized Water Reactor (PWR) Marine Nuclear Power Plant (MNPP). (authors)

  17. Analysis to verify effectiveness of alternative cooling method in case of loss of RHR function during mid-loop operation

    International Nuclear Information System (INIS)

    Nagae, Takashi; Tamaki, Tomohiko; Murase, Michio; Ayano, Teruyoshi

    2003-01-01

    In the mid-loop operation during shutdown of the pressurized water reactor (PWR) plant, the core decay heat is cooled by the residual heat removal (RHR) system. In the case of loss of the RHR function, core cooling is achieved by reflux cooling through the steam generator (SG) when the reactor coolant system (RCS) is closed, or by gravity injection of water from the refueling water storage pit (RWSP) when a large opening is present in the RCS. However, it is uncertain whether core cooling can be achieved by these alternative cooling methods, if the opening is not large enough in the RCS. In this study, the effectiveness of the reflux cooling through the SG and the gravity injection of water from the RWSP in the mid-loop operation three days after shutdown was investigated by using RELAP5/MOD3.2 with a plant model representing a typical 4-loop PWR plant in Japan, assuming that two bases of the pressurizer safety valves were removed. As a result, it was verified that in the case of a combination of the reflux cooling by through the SG and gravity injection of water from the RWSP, the time until the core was uncovered with water extended about an hour from that in the case of no cooling method. (author)

  18. 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)

  19. Implementation of an RHR/LPSI pump coupling retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; Koch, R.P.; Orewyler, R.; Tipton, J.W.

    1994-01-01

    Nuclear plant operating experience has shown the RHR and LPSI services to be very demanding on pumps. The systems handle borated water at high temperatures and pressures with frequent step changes in both temperature and pressure. Additionally, the industry trend towards reduced flow rates during plant mid-loop (reduced inventory) conditions has resulted in extended pump operation at flow rates significantly below the pump best efficiency point flow. Operation at these low flow fates is known to cause high thrust loads and large shaft deflections. The combination of these and other factors have resulted in short mechanical seal life and short motor bearing life, thus requiring frequent pump and motor maintenance. For many nuclear plants, including Southern California Edison's (SCE) San Onofre Units 2 and 3, these pumps have represented a major operations and maintenance (O ampersand M) expenditure and a significant source of radiation exposure to plant personnel. SCE management determined that a pump upgrade was justified to reduce the O ampersand M costs and to improve plant availability. SCE decided to proceed with a pump retrofit program to improve the pump maintainability, reliability and availability. Installation was completed for four LPSI pumps at San Onofre Units 2 and 3 during the Cycle 7 refueling outages in 1993. A key to the program's success was the removal of many traditional supplier and customer barriers and revision of supplier and customer roles to create a unified team. This paper traces the RHR/LPSI retrofit program for San Onofre from problem identification to project implementation. The team approach used for this program and the lessons learned may be useful to other utilities and vendors when evaluating or implementing system and equipment upgrades

  20. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  1. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  2. Regulatory analysis for the resolution of Generic Issue 99: Loss of RHR [residual heat removal] capability in PWRs

    International Nuclear Information System (INIS)

    Spano, A.H.

    1989-02-01

    Generic Issue 99 is concerned with the loss of residual heat removal (RHR) capability in pressurized water reactors during cold-plant outage operations. The issue focuses on two risk-significant common-cause failure modes of the RHR system: (1) air binding of the RHR pumps during reduced-inventory operations and (2) spurious closure of the RHR suction valves due to misapplication of the autoclosure interlocks. Resolution of this issue involves consideration of the adequacy of plant capabilities for (1) preventing losses of RHR, (2) responding promptly and effectively to such challenges in order to prevent core damage, and (3) ensuring timely containment protection against the release of radioactivity to the environment in the unlikely event of core damage due to loss of shutdown cooling. This entails examination of (1) relevant operational and accident response procedures, (2) the instrumentation available to the operator for accident diagnosis and mitigation, and (3) the administrative controls available for ensuring control room cognizance of ongoing maintenance activities that could potentially affect the stability of the reactor coolant system. This regulatory analysis provides quantitative assessments of the costs and benefits associated with several alternatives considered for the resolution of Generic Issue 99. 24 refs

  3. NPP Krsko analysis of the loss of the RHR system during mid-loop operation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Debrecin, N.

    2002-01-01

    In the paper the results of the NPP Krsko analysis of the loss of the Residual Heat Removal (RHR) system at mid-loop conditions using RELAP5/MOD 3.2.2 Gamma and RELAP5/MOD 3.3 Beta code are presented. Both pressurizer and Steam Generator (SG) 1 manway were open. The facility was open to the containment atmosphere and filled with air above liquid level. Loss of the RHR system when Reactor Coolant System (RCS) is open causes quick boiling in the core and loss of the inventory available for the cooling of the core through the openings. Aims of the analysis were threefold. First, the consequences of the transient for NPP Krsko, i.e., the time to core uncovery was determined. Secondly, the influence of the applied RELAP5 code version (MOD 3.2.2 Gamma and MOD 3.3 Beta) in the analysis of that particular transient with noncondensable gases was assessed. Third, the analysis with SG secondary sides under wet lay-up conditions was performed in order to assess the influence of condensation in the SG U tubes on liquid inventory in the system and core cooling capability.(author)

  4. GnRHR-II knockdown swine have constitutively lower serum testosterone concentrations, impaired senstitivity to GnRH analogues and reduced semen quality

    Science.gov (United States)

    The second mammalian GnRH isoform (GnRH-II) and its specific receptor (GnRHR-II) are abundantly produced within swine testes. GnRHR-II localizes to porcine Leydig cells and exogenous GnRH-II treatment robustly stimulates testosterone production in vivo, despite minimal secretion of luteinizing hormo...

  5. 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.

  6. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  7. RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Bajs, T.; Prah, M.

    1998-01-01

    In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

  8. Improvement on models associated with LOCA and loss of RHR accidents during shutdown

    International Nuclear Information System (INIS)

    Chang, W. P.; Chung, Y. J.; Kim, W. S.; Kim, K. D.; Lee, S. J.; Jung, J. J.; Ha, G. S.; Son, Y. S.; Chung, B. D.; Han, D. H.; Lee, Y. J.; Hwang, T. S.; Lee, S. Y.; Park, C. Y.; Choi, H. R.; Lee, S. Y.; Choi, J. H.; Ban, C. H.; Bae, G. H.

    1997-07-01

    The characteristics of the best estimate codes available in Korea have been studied through literature surveys for the reliability on LOCA analyses and then, a feasibility study on reduction of capacities of existing safety systems in YGN 3/4 have been carried out using the codes. Since it has been expected to adopt DVI + 4 -Train HPSI in the next generation reactor, the core uncoveries under one DVI line break and 6 cold leg break, which is a requirement for advance d reactor by EPRI, in addition to LBLOCA for reduction effect of SIT capacity, have been analyzed. Finally, an effort on finding the way how the system could be simplified, has been made through the analysis of SIT injection characteristics. On the other hand, the best estimate methodology consisting of uncertainties of the code itself, bias, and application have been developed first and quantification of the uncertainty has been made the case of KORI unit 3 afterward. The prediction capabilities of the best estimate codes and major physical models on the accident under loss of RHR during shutdown have been assessed suing the large scale experimental data delivered from France and then, the assessed codes have been used to provide essential data required for description of operation procedures in YGN 3/4. (author). 64 refs., 45 figs

  9. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  10. Ammonia excretion in Caenorhabditis elegans: mechanism and evidence of ammonia transport of the Rhesus protein CeRhr-1

    Science.gov (United States)

    Adlimoghaddam, Aida; Boeckstaens, Mélanie; Marini, Anna-Maria; Treberg, Jason R.; Brassinga, Ann-Karen C.; Weihrauch, Dirk

    2015-01-01

    ABSTRACT The soil-dwelling nematode Caenorhabditis elegans is a bacteriovorous animal, excreting the vast majority of its nitrogenous waste as ammonia (25.3±1.2 µmol gFW−1 day−1) and very little urea (0.21±0.004 µmol gFW−1 day−1). Although these roundworms have been used for decades as genetic model systems, very little is known about their strategy to eliminate the toxic waste product ammonia from their bodies into the environment. The current study provides evidence that ammonia is at least partially excreted via the hypodermis. Starvation reduced the ammonia excretion rates by more than half, whereas mRNA expression levels of the Rhesus protein CeRhr-2, V-type H+-ATPase (subunit A) and Na+/K+-ATPase (α-subunit) decreased correspondingly. Moreover, ammonia excretion rates were enhanced in media buffered to pH 5 and decreased at pH 9.5. Inhibitor experiments, combined with enzyme activity measurements and mRNA expression analyses, further suggested that the excretion mechanism involves the participation of the V-type H+-ATPase, carbonic anhydrase, Na+/K+-ATPase, and a functional microtubule network. These findings indicate that ammonia is excreted, not only by apical ammonia trapping, but also via vesicular transport and exocytosis. Exposure to 1 mmol l−1 NH4Cl caused a 10-fold increase in body ammonia and a tripling of ammonia excretion rates. Gene expression levels of CeRhr-1 and CeRhr-2, V-ATPase and Na+/K+-ATPase also increased significantly in response to 1 mmol l−1 NH4Cl. Importantly, a functional expression analysis showed, for the first time, ammonia transport capabilities for CeRhr-1 in a phylogenetically ancient invertebrate system, identifying these proteins as potential functional precursors to the vertebrate ammonia-transporting Rh-glycoproteins. PMID:25740900

  11. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  12. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  13. On-line filtering

    International Nuclear Information System (INIS)

    Verkerk, C.

    1978-01-01

    Present day electronic detectors used in high energy physics make it possible to obtain high event rates and it is likely that future experiments will face even higher data rates than at present. The complexity of the apparatus increases very rapidly with time and also the criteria for selecting desired events become more and more complex. So complex in fact that the fast trigger system cannot be designed to fully cope with it. The interesting events become thus contaminated with multitudes of uninteresting ones. To distinguish the 'good' events from the often overwhelming background of other events one has to resort to computing techniques. Normally this selection is made in the first part of the analysis of the events, analysis normally performed on a powerful scientific computer. This implies however that many uninteresting or background events have to be recorded during the experiment for subsequent analysis. A number of undesired consequences result; and these constitute a sufficient reason for trying to perform the selection at an earlier stage, in fact ideally before the events are recorded on magnetic tape. This early selection is called 'on-line filtering' and it is the topic of the present lectures. (Auth.)

  14. On line portal imaging

    International Nuclear Information System (INIS)

    Munro, Peter

    1997-01-01

    Purpose/Objective: The purpose of this presentation is to examine the various imaging devices that have been developed for portal imaging, describe some of the image registration methods that have been developed to determine geometric errors quantitatively, and discuss how portal imaging has been incorporated into clinical practice. Discussion: Verification of patient positioning has always been an important aspect of external beam radiation therapy. Over the past decade many portal imaging devices have been developed by individual investigators and most accelerator manufacturers now offer 'on-line' portal imaging systems. The commercial devices include T.V. camera-based systems, liquid ionisation chamber systems, and shortly, flat panel systems. The characteristics of these imaging systems will be discussed. In addition, other approaches such as the use of kilovoltage x-ray sources, video monitoring, and ultrasound have been proposed for improving patient positioning. Some of the advantages of these approaches will be discussed. One of the major advantages of on-line portal imaging is that many quantitative techniques have been developed to detect errors in patient positioning. The general approach is to register anatomic structures on a portal image with the same structures on a digitized simulator film. Once the anatomic structures have been registered, any discrepancies in the position of the patient can be identified. One problem is finding a common frame of reference for the simulator and portal images, since the location of the radiation field within the pixel matrix may differ for the two images. As a result, a common frame of reference has to be established before the anatomic structures in the images can be registered - generally by registering radiation field edges identified in the simulator and portal images. In addition, distortions in patient geometry or rotations out of the image plane can confound the image registration techniques. Despite the

  15. On line portal imaging

    International Nuclear Information System (INIS)

    Munro, Peter

    1996-01-01

    Purpose/Objective: The purpose of this presentation is to examine the various imaging devices that have been developed for portal imaging; describe some of the image registration methods that have been developed to determine geometric errors quantitatively; discuss some of the ways that portal imaging has been incorporated into routine clinical practice; describe quality assurance procedures for these devices, and discuss the use of portal imaging devices for dosimetry applications. Discussion: Verification of patient positioning has always been an important aspect of external beam radiation therapy. Over the past decade many portal imaging devices have been developed by individual investigators and most accelerator manufacturers now offer 'on-line' portal imaging systems. The commercial devices can be classified into three categories: T.V. camera-based systems, liquid ionisation chamber systems, and amorphous silicon systems. Many factors influence the quality of images generated by these portal imaging systems. These include factors which are unavoidable (e.g., low subject contrast), factors which depend upon the individual imaging device forming the image (e.g., dose utilisation, spatial resolution) as well as factors which depend upon the characteristics of the linear accelerator irradiating the imaging system (x-ray source size, image magnification). The characteristics of individual imaging systems, such as spatial resolution, temporal response, and quantum utilisation will be discussed. One of the major advantages of on-line portal imaging is that many quantitative techniques have been developed to detect errors in patient positioning. The general approach is to register anatomic structures on a portal image with the same structures on a digitized simulator film. Once the anatomic structures have been registered, any discrepancies in the position of the patient can be identified. However, the task is not nearly as straight-forward as it sounds. One problem

  16. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  17. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  18. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  19. On line portal imaging

    International Nuclear Information System (INIS)

    Munro, Peter

    1995-01-01

    Purpose/Objective: The purpose of this presentation is to review the physics of imaging with high energy x-ray beams; examine the various imaging devices that have been developed for portal imaging; describe some of the image registration methods that have been developed to determine errors in patient positioning quantitatively; and discuss some of the ways that portal imaging has been incorporated into routine clinical practice. Verification of patient positioning has always been an important aspect of external beam radiation therapy. Checks of patient positioning have generally been done with film, however, film suffers from a number of drawbacks, such as poor image display and delays due to film development. Over the past decade many portal imaging devices have been developed by individual investigators and most accelerator manufacturers now offer 'on-line' portal imaging systems, which are intended to overcome the limitations of portal films. The commercial devices can be classified into three categories: T.V. camera-based systems, liquid ionisation chamber systems, and amorphous silicon systems. Many factors influence the quality of images generated by these portal imaging systems. These include factors which are unavoidable (e.g., low subject contrast), factors which depend upon the individual imaging device forming the image (e.g., dose utilisation, spatial resolution) as well as factors which depend upon the characteristics of the linear accelerator irradiating the imaging system (x-ray source size, image magnification). The fundamental factors which limit image quality and the characteristics of individual imaging systems, such as spatial resolution, temporal response, and quantum utilisation will be discussed. One of the major advantages of on-line portal imaging is that many quantitative techniques have been developed to detect errors in patient positioning. The general approach is to register anatomic structures on a portal image with the same

  20. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  1. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  2. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  3. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  4. On-line fatigue monitoring and margins probabilistic assessment

    International Nuclear Information System (INIS)

    Fournier, I.; Morilhat, P.

    1993-01-01

    An on-line computer aided system has been developed by Electricite de France, the French utility, for a fatigue monitoring of critical locations in the nuclear steam supply system. This tool, called fatiguemeter, includes as input data only existing plant parameters and is based on some conservative assumptions at several steps of the damage assessment (thermal boundary conditions, stress computation...). This paper presents recent developments performed toward a better assessing of margins involved in the complete analysis. The methodology is enlightened with an example showing the influence of plant parameters incertitude on the final stress computed at a PWR 900 MW unit pressurizer surge line nozzle. (author)

  5. Biosynthesis of gonadotropin-releasing hormone (GnRH) and GnRH receptor (GnRHR) in hypothalamic-pituitary unit of anoestrous and cyclic ewes.

    Science.gov (United States)

    Ciechanowska, M O; Łapot, M; Mateusiak, K; Paruszewska, E; Malewski, T; Przekop, F

    2017-02-01

    This study was performed to explain how the molecular processes governing the biosynthesis of gonadotropin-releasing hormone (GnRH) and GnRH receptor (GnRHR) in the hypothalamic-pituitary unit are reflected by luteinizing hormone (LH) secretion in sheep during anoestrous period and during luteal and follicular phases of the oestrous cycle. Using an enzyme-linked immunosorbent assay (ELISA), we analyzed the levels of GnRH and GnRHR in preoptic area (POA), anterior (AH) and ventromedial hypothalamus (VM), stalk-median eminence (SME), and GnRHR in the anterior pituitary gland (AP). Radioimmunoassay has also been used to define changes in plasma LH concentrations. The study provides evidence that the levels of GnRH in the whole hypothalamus of anoestrous ewes were lower than that in sheep during the follicular phase of the oestrous cycle (POA: p pituitary unit, as well as LH level, in the blood in anoestrous ewes were significantly lower than those detected in animals of both cyclic groups. Our data suggest that decrease in LH secretion during the long photoperiod in sheep may be due to low translational activity of genes encoding both GnRH and GnRHR.

  6. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  7. The study of gravity makeup to RCS for the loss of RHR event during mid-loop operation

    International Nuclear Information System (INIS)

    Oh, H. S.; Yoon, D. J.; Ha, S. J.; Lee, C. S.

    2004-01-01

    In case of the loss of residual heat removal system (RHR) event during mid-loop operation, one of the mitigation actions to prevent core uncovery is gravity makeup to the RCS. This study includes the mitigation actions for gravity makeup to the RCS for 3-loop nuclear power plant, minimum gravity makeup flow for prevention of core boiling and core uncovery and possible pass of gravity make up. Also, the evaluation of minimum gravity makeup to prevent core boiling and core uncovery was performed using the RELAP/MOD3.2.2beta code. The results of this study show that the minimum flow to prevent core uncovery in case of cold leg injection (about 20m 3 /hr) is too small to recover the core water level. So, our conclusion is that the minimum flow to prevent core boiling (about 170m 3 /hr) is enough to recover core water level

  8. 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

  9. PWR surveillance based on correspondence between empirical models and physical

    International Nuclear Information System (INIS)

    Zwingelstein, G.; Upadhyaya, B.R.; Kerlin, T.W.

    1976-01-01

    An on line surveillance method based on the correspondence between empirical models and physicals models is proposed for pressurized water reactors. Two types of empirical models are considered as well as the mathematical models defining the correspondence between the physical and empirical parameters. The efficiency of this method is illustrated for the surveillance of the Doppler coefficient for Oconee I (an 886 MWe PWR) [fr

  10. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  11. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  12. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  13. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  14. Aberrant gonadotropin-releasing hormone receptor (GnRHR) expression and its regulation of CYP11B2 expression and aldosterone production in adrenal aldosterone-producing adenoma (APA).

    Science.gov (United States)

    Nakamura, Yasuhiro; Hattangady, Namita G; Ye, Ping; Satoh, Fumitoshi; Morimoto, Ryo; Ito-Saito, Takako; Sugawara, Akira; Ohba, Koji; Takahashi, Kazuhiro; Rainey, William E; Sasano, Hironobu

    2014-03-25

    Aberrant expression of gonadotropin-releasing hormone receptor (GnRHR) has been reported in human adrenal tissues including aldosterone-producing adenoma (APA). However, the details of its expression and functional role in adrenals are still not clear. In this study, quantitative RT-PCR analysis revealed the mean level of GnRHR mRNA was significantly higher in APAs than in human normal adrenal (NA) (P=0.004). GnRHR protein expression was detected in human NA and neoplastic adrenal tissues. In H295R cells transfected with GnRHR, treatment with GnRH resulted in a concentration-dependent increase in CYP11B2 reporter activity. Chronic activation of GnRHR with GnRH (100nM), in a cell line with doxycycline-inducible GnRHR (H295R-TR/GnRHR), increased CYP11B2 expression and aldosterone production. These agonistic effects were inhibited by blockers for the calcium signaling pathway, KN93 and calmidazolium. These results suggest GnRH, through heterotopic expression of its receptor, may be a potential regulator of CYP11B2 expression levels in some cases of APA. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  15. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  16. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  17. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  18. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  19. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  20. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  1. On-line method to identify control rod drops in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Souza, T.J.; Martinez, A.S.; Medeiros, J.A.C.C.; Palma, D.A.P.; Gonçalves, A.C.

    2014-01-01

    Highlights: • On-line method to identify control rod drops in PWR reactors. • Identification method based on the readings of the ex-core detector. • Recognition of the patterns in the ex-core detector responses. - Abstract: A control rod drop event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimise undesirable effects in such a scenario. The goal of this work is to develop an online method to identify control rod drops in PWR reactors. The method entails the construction of a tool based on ex-core detector responses. It proposes to recognize patterns in the neutron ex-core detectors responses and thus to make an online identification of a control rod drop in the core during the reactor operation. The results of the study, as well as the behaviour of the detector responses demonstrated the feasibility of this method

  2. Alterations to dendritic spine morphology, but not dendrite patterning, of cortical projection neurons in Tc1 and Ts1Rhr mouse models of Down syndrome.

    Directory of Open Access Journals (Sweden)

    Matilda A Haas

    Full Text Available Down Syndrome (DS is a highly prevalent developmental disorder, affecting 1/700 births. Intellectual disability, which affects learning and memory, is present in all cases and is reflected by below average IQ. We sought to determine whether defective morphology and connectivity in neurons of the cerebral cortex may underlie the cognitive deficits that have been described in two mouse models of DS, the Tc1 and Ts1Rhr mouse lines. We utilised in utero electroporation to label a cohort of future upper layer projection neurons in the cerebral cortex of developing mouse embryos with GFP, and then examined neuronal positioning and morphology in early adulthood, which revealed no alterations in cortical layer position or morphology in either Tc1 or Ts1Rhr mouse cortex. The number of dendrites, as well as dendrite length and branching was normal in both DS models, compared with wildtype controls. The sites of projection neuron synaptic inputs, dendritic spines, were analysed in Tc1 and Ts1Rhr cortex at three weeks and three months after birth, and significant changes in spine morphology were observed in both mouse lines. Ts1Rhr mice had significantly fewer thin spines at three weeks of age. At three months of age Tc1 mice had significantly fewer mushroom spines--the morphology associated with established synaptic inputs and learning and memory. The decrease in mushroom spines was accompanied by a significant increase in the number of stubby spines. This data suggests that dendritic spine abnormalities may be a more important contributor to cognitive deficits in DS models, rather than overall neuronal architecture defects.

  3. CERN Video News on line

    CERN Multimedia

    2003-01-01

    The latest CERN video news is on line. In this issue : an interview with the Director General and reports on the new home for the DELPHI barrel and the CERN firemen's spectacular training programme. There's also a vintage video news clip from 1954. See: www.cern.ch/video or Bulletin web page

  4. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  5. On-line plant-wide monitoring using neural networks

    International Nuclear Information System (INIS)

    Turkcan, E.; Ciftcioglu, O.; Eryurek, E.; Upadhyaya, B.R.

    1992-06-01

    The on-line signal analysis system designed for a multi-level mode operation using neural networks is described. The system is capable of monitoring the plant states by tracking different number of signals up to 32 simultaneously. The data used for this study were acquired from the Borssele Nuclear Power Plant (PWR type), and using the on-line monitoring system. An on-line plant-wide monitoring study using a multilayer neural network model is discussed in this paper. The back-propagation neural network algorithm is used for training the network. The technique assumes that each physical state of the power plant can be represented by a unique pattern of instrument readings which can be related to the condition of the plant. When disturbance occurs, the sensor readings undergo a transient, and form a different set of patterns which represent the new operational status. Diagnosing these patterns can be helpful in identifying this new state of the power plant. To this end, plant-wide monitoring with neutral networks is one of the new techniques in real-time applications. (author). 9 refs.; 5 figs

  6. On-line moisture analysis

    CERN Document Server

    Cutmore, N G

    2002-01-01

    Measurement of the moisture content of iron ore has become a key issue for controlling moisture additions for dust suppression. In most cases moisture content is still determined by manual or automatic sampling of the ore stream, followed by conventional laboratory analysis by oven drying. Although this procedure enables the moisture content to be routinely monitored, it is too slow for control purposes. This has generated renewed interest in on-line techniques for the accurate and rapid measurement of moisture in iron ore on conveyors. Microwave transmission techniques have emerged over the past 40 years as the dominant technology for on-line measurement of moisture in bulk materials, including iron ores. Alternative technologies have their limitations. Infra-red analysers are used in a variety of process industries, but rely on the measurement of absorption by moisture in a very thin surface layer. Consequently such probes may be compromised by particle size effects and biased presentation of the bulk mater...

  7. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  8. Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

    Energy Technology Data Exchange (ETDEWEB)

    Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

    2016-04-15

    Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

  9. On line ultrasonic integrated backscatter

    International Nuclear Information System (INIS)

    Landini, L.; Picano, E.; Mazzarisi, A.; Santarelli, F.; Benassi, A.; De Pieri, G.

    1988-01-01

    A new equipment for on-line evaluation of index based on two-dimensional integrated backscatter from ultrasonic images is described. The new equipment is fully integrated into a B-mode ultrasonic apparatus which provides a simultaneous display of conventional information together with parameters of tissue characterization. The system has been tested with a backscattering model of microbubbles in polysaccharide solution, characterized by a physiological exponential time decay. An exponential fitting to the experimental data was performed which yielded r=0.95

  10. On-line nuclear orientation

    International Nuclear Information System (INIS)

    Krane, K.S.

    1990-01-01

    This grant has as its overall goal the pursuit of on-line nuclear orientation experiments for the purpose of eliciting details of nuclear structure from the decays of neutron-deficient nuclei, such as those produced by the Holifield Heavy-Ion Research Facility at Oak Ridge and extracted by the UNISOR Isotope Separator. This paper discusses: refrigerator development; the decay of 184 Au; the decay of 191 Hg to 191 Au; the decay of 189 Pt to 189 Ir; the decays of 109,111 Pd; the decay of 172 Er; and solid angle corrections

  11. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  12. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  13. On-line measurements of RuO{sub 4} during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  14. On-line moisture analysis

    International Nuclear Information System (INIS)

    Cutmore, N.G.; Mijak, D.G

    2002-01-01

    Measurement of the moisture content of iron ore has become a key issue for controlling moisture additions for dust suppression. In most cases moisture content is still determined by manual or automatic sampling of the ore stream, followed by conventional laboratory analysis by oven drying. Although this procedure enables the moisture content to be routinely monitored, it is too slow for control purposes. This has generated renewed interest in on-line techniques for the accurate and rapid measurement of moisture in iron ore on conveyors. Microwave transmission techniques have emerged over the past 40 years as the dominant technology for on-line measurement of moisture in bulk materials, including iron ores. Alternative technologies have their limitations. Infra-red analysers are used in a variety of process industries, but rely on the measurement of absorption by moisture in a very thin surface layer. Consequently such probes may be compromised by particle size effects and biased presentation of the bulk material. Nuclear-based analysers measure the total hydrogen content in the sample and do not differentiate between free and combined moisture. Such analysers may also be sensitive to material presentation and elemental composition. Very low frequency electromagnetic probes, such as capacitance or conductance probes, operate in the frequency region where the DC conductivity dominates much of the response, which is a function not only of moisture content but also of ionic composition and chemistry. These problems are overcome using microwave transmission techniques, which also have the following advantages, as a true bulk moisture analysis is obtained, because a high percentage of the bulk material is analysed; the moisture estimate is mostly insensitive to any biased presentation of moisture, for example due to stratification of bulk material with different moisture content and because no physical contact is made between the sensor and the bulk material. This is

  15. Effect of endotoxin on the expression of GnRH and GnRHR genes in the hypothalamus and anterior pituitary gland of anestrous ewes.

    Science.gov (United States)

    Herman, Andrzej Przemysław; Tomaszewska-Zaremba, Dorota

    2010-07-01

    An immune/inflammatory challenge can affect reproduction at the level of the hypothalamus, pituitary gland, or gonads. Nonetheless, the major impact is thought to occur within the brain or the pituitary gland. The present study was designed to examine the effect of intravenous (i.v.) lipopolysaccharide (LPS) injection on the expression of gonadotropin-releasing hormone (GnRH) and the gonadotropin-releasing hormone receptor (GnRHR) genes in the hypothalamic structures where GnRH neurons are located as well as in the anterior pituitary gland (AP) of anestrous ewes. We also determined the effect of LPS on luteinizing hormone (LH) release. It was found that i.v. LPS injection significantly decreased GnRH and GnRHR mRNAs levels in the preoptic area (40%, ppituitary cells to GnRH stimulation. The presence of GnRH mRNA in the median eminence, the hypothalamic structure where GnRH-ergic neurons' terminals are located, suggests that the axonal transport of GnRH mRNA may occur in these neurons. This phenomenon could play an important role in the physiology of GnRH neurons. Our data demonstrate that immune stress could be important inhibitor of this process. Copyright 2010 Elsevier B.V. All rights reserved.

  16. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  17. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  18. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  19. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  20. On-line data display

    Science.gov (United States)

    Lang, Sherman Y. T.; Brooks, Martin; Gauthier, Marc; Wein, Marceli

    1993-05-01

    A data display system for embedded realtime systems has been developed for use as an operator's user interface and debugging tool. The motivation for development of the On-Line Data Display (ODD) have come from several sources. In particular the design reflects the needs of researchers developing an experimental mobile robot within our laboratory. A proliferation of specialized user interfaces revealed a need for a flexible communications and graphical data display system. At the same time the system had to be readily extensible for arbitrary graphical display formats which would be required for data visualization needs of the researchers. The system defines a communication protocol transmitting 'datagrams' between tasks executing on the realtime system and virtual devices displaying the data in a meaningful way on a graphical workstation. The communication protocol multiplexes logical channels on a single data stream. The current implementation consists of a server for the Harmony realtime operating system and an application written for the Macintosh computer. Flexibility requirements resulted in a highly modular server design, and a layered modular object- oriented design for the Macintosh part of the system. Users assign data types to specific channels at run time. Then devices are instantiated by the user and connected to channels to receive datagrams. The current suite of device types do not provide enough functionality for most users' specialized needs. Instead the system design allows the creation of new device types with modest programming effort. The protocol, design and use of the system are discussed.

  1. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  2. Study for identification of control rod drops in PWR reactors at any burnup step

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2013-01-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  3. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  4. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  5. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  6. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  7. On-Line Monitoring of Instrument Channel Performance in Nuclear Power Plant Using PEANO

    International Nuclear Information System (INIS)

    Fantoni, Paolo F.; Hoffmann, Mario; Shankar, Ramesh; Davis, Eddie L.

    2002-01-01

    On-Line monitoring evaluates instrument channel performance by assessing its consistency with other plant indications. Industry and EPRI experience at several plants has shown this overall approach to be very effective in identifying instrument channels that are exhibiting degrading or inconsistent performance characteristics. On-Line monitoring of instrument channels provides information about the condition of the monitored channels through accurate, more frequent monitoring of each channel's performance over time. This type of performance monitoring is a methodology that offers an alternate approach to traditional time-directed calibration. On-line monitoring of these channels can provide an assessment of instrument performance and provide a basis for determining when adjustments are necessary. Elimination or reduction of unnecessary field calibrations can reduce associated labor costs, reduce personnel radiation exposure and reduce the potential for miss-calibration. PEANO is a system for on-line calibration monitoring developed in the years 1995-2000 at the Institutt for energiteknikk (IFE), Norway, which makes use of Artificial Intelligence techniques for its purpose. The system has been tested successfully in Europe in off-line tests with EDF (France), Tecnatom (Spain) and ENEA (Italy). PEANO is currently installed and used for on-line monitoring at the HBWR reactor in Halden. This paper describes the results of performance tests on PEANO with real data from a US PWR plant, in the framework of a co-operation among IFE, EPRI and Edan Engineering, to evaluate the potentials of PEANO for future installations in US nuclear plants. (authors)

  8. Electrochemical measurements in PWR steam generators to follow crevice chemistry

    International Nuclear Information System (INIS)

    Feron, D.

    1991-01-01

    In PWR steam generator crevices, the evolution of chemistry is important for the understanding of corrosion phenomena. Electrochemical measurements have been performed in high temperature simulated crevice environments in order to follow hideout processes and remedial actions (on-line addition of boric acid). Reported tests have been conducted with model boilers of AJAX facilities. Eccentric and concentric tube support plate crevices have been instrumented with platinum electrodes. Electrochemical measurements have been collected when model boiler was under nominal conditions (primary temperature: 335 deg C, secondary temperature: 280 deg C). They include Electrochemical Impedance Spectroscopy (EIS) and potential measurements: with EIS, sodium and boric acid hideouts have been detected and followed. Potential measurements have been performed in an attempt to measure crevice PH evolution

  9. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  10. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  11. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  12. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  13. A development of maintenance educational support method by using navigation method. Pt. 3. A detal design of RHR pump maintenance educational support prototype system

    International Nuclear Information System (INIS)

    Yoshino, Kenji; Hirotsu, Yuko; Fujimoto, Junzo; Tsumura, Joji

    2003-01-01

    This research establishes the new maintenance educational support prototype system for aiming at organization as Trinity (a supervision person, a construction person in charge, a work person in charge, and group length), and improvement in human reliability for the purpose of aiming at before hand prevention of a systematic error, human error, and the accident and a trouble. Main results: (1) This paper prototype system acquired a participant data by the input of 'position and rank' etc. of a participant, and was taken as the structure which navigates the participant exactly from the start of RHR pump maintenance work to the end on the basis of the data (question etc.) while performing reeducation based on the reply of an organization, organization, and the participant to the structure-ized question in connection with on-site work, judging, and the judgment result etc. by the system. (2) RHR pump maintenance work has the complicated hierarchy. Then, in order to raise the study effect of the participant about maintenance education, it considered as the structure as which that a participant does what study, further to which process a participant next progresses in which protion of the whole maintenance process a participant is and etc. A participant can always keep regarding the whole study in connection with maintenance education (Fig. 2). (3) This paper prototype system was taken as the structure which shows clearly a participant special feature, weak point, etc., by plotting an organization, organization, and the reply result of the subject of the question completion before in connection with on-site work, and after completion on the same screen. (4) This paper prototype system Judge how changed a participant action pattern's completion front, and after completion. For example, from the organization neglect/duty neglect type before completion, after completion specifies having changed to the organization/duty serious-consideration type, further, is adding the

  14. A development of maintenance educational support method by using navigation method. Pt. 1. A conceptual design of RHR pump maintenance educational support prototype system

    International Nuclear Information System (INIS)

    Yoshino, Kenji; Hirotsu, Yuko; Fujimoto, Junzo; Gouda, Hidenori

    2001-01-01

    The purpose of this research is before hand prevention of the systematic error produced from the fault of command / command system on the different organization in a nuclear power plant, and the individual error of the demand level of maintenance work, and knowledge and experience level of a maintenance worker produced from it being incongruent (mismatch), and generating of the serious accident and a serious trouble. This research attains optimization with the difficulty of maintenance work, and work execution capability, such as the persons concerned (a supervision person, a construction person, work person), and proposes the new 'maintenance educational support technique' which is useful to the improvement in 'work reliability' and 'before hand prevention of a human error' in the maintenance work spot. The main results concerning this research are shown below. (1) In this research, concept design of RHR pump maintenance work educational support prototype system was performed on the basis of the examination result of a process analysis of RHR pump maintenance work, load mapping, a rule base/knowledge base and a data base/filing system. Consequently, when an educator-ed talked with a data base/filing system along with load mapping of maintenance work (Why, How, What), a possibility that it could reach exactly from work start to the sub goal or the last goal and timely was suggested. (2) As a result of classifying 'the difficulty of maintenance work', and man's work execution capability' and comparing the result using SRK (behavior of human) index which Rasmussen proposed, among both, it turns out that difference (inharmonious) arises. The exact presentation of the educational method or educational material which responded to knowledge and experience level of an educational achievement target or an educator-ed can be referred to as being connected with the improvement in work reliability, and before hand prevention of a human error as a result by then, incorporating

  15. On-line detection of key radionuclides for fuel-rod failure in a pressurized water reactor.

    Science.gov (United States)

    Qin, Guoxiu; Chen, Xilin; Guo, Xiaoqing; Ni, Ning

    2016-08-01

    For early on-line detection of fuel rod failure, the key radionuclides useful in monitoring must leak easily from failing rods. Yield, half-life, and mass share of fission products that enter the primary coolant also need to be considered in on-line analyses. From all the nuclides that enter the primary coolant during fuel-rod failure, (135)Xe and (88)Kr were ultimately chosen as crucial for on-line monitoring of fuel-rod failure. A monitoring system for fuel-rod failure detection for pressurized water reactor (PWR) based on the LaBr3(Ce) detector was assembled and tested. The samples of coolant from the PWR were measured using the system as well as a HPGe γ-ray spectrometer. A comparison showed the method was feasible. Finally, the γ-ray spectra of primary coolant were measured under normal operations and during fuel-rod failure. The two peaks of (135)Xe (249.8keV) and (88)Kr (2392.1keV) were visible, confirming that the method is capable of monitoring fuel-rod failure on-line. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. Shawnee Mission's On-Line Cataloging System

    Directory of Open Access Journals (Sweden)

    Ellen Wasby Miller

    1971-03-01

    Full Text Available An on-line cataloging pilot project for two elementary schools is discussed. The system components are 2740 terminals, upper-lower-case input, IBM's FASTER generalized software package, and usual cards/labels output. Reasons for choosing FASTER, software and hardware features, operating procedures, system performance and costs are detailed. Future expansion to cataloging 100,000 annual K-12 acquisitions, on-line circulation, retrospective conversion, and union book catalogs is set forth.

  17. INIS retrieval service, towards on-line

    International Nuclear Information System (INIS)

    Ebinuma, Yukio; Komatsubara, Yasutoshi

    1983-01-01

    Japan Atomic Energy Research Institute executes the retrieval service of INIS atomic energy information by batch system in cooperation with Genshiryoku Kozaikai. This service is very popular to the users in whole Japan, but the demand of on-line service has increased recently. Therefore, it was decided to begin the INIS on-line service from January, 1984, through the on-line information retrieval system of the Japan Information Center of Science and Technology. It is expected that when the operation will be started, the utilization of INIS atomic energy information in Japan will drastically increase. Also Japan Atomic Energy Research Institute has carried out the retrieval service by on-line system for those in the institute besides the batch system, accordingly, at this opportunity, the state of utilization of both systems and their distinction to use effectively, and the operation and the method of utilization of the on-line information retrieval system of JICST are explained. In the on-line system, the users are accessible to the data base themselves, and immediate information retrieval is possible, while in the batch system, the related information can be retrieved without fail, and the troublesome operation of equipment is not necessary. (Kako, I.)

  18. Model-based fault detection and isolation of a PWR nuclear power plant using neural networks

    International Nuclear Information System (INIS)

    Far, R.R.; Davilu, H.; Lucas, C.

    2008-01-01

    The proper and timely fault detection and isolation of industrial plant is of premier importance to guarantee the safe and reliable operation of industrial plants. The paper presents application of a neural networks-based scheme for fault detection and isolation, for the pressurizer of a PWR nuclear power plant. The scheme is constituted by 2 components: residual generation and fault isolation. The first component generates residuals via the discrepancy between measurements coming from the plant and a nominal model. The neutral network estimator is trained with healthy data collected from a full-scale simulator. For the second component detection thresholds are used to encode the residuals as bipolar vectors which represent fault patterns. These patterns are stored in an associative memory based on a recurrent neutral network. The proposed fault diagnosis tool is evaluated on-line via a full-scale simulator detected and isolate the main faults appearing in the pressurizer of a PWR. (orig.)

  19. Current status of low power/shutdown PSA and accident sequence analysis for loss of RHR during mid-loop operation

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Choi, Young; Kim, Tae Woon; Jin, Young Ho

    1994-07-01

    Probabilistic safety assessment (PSA) has been applied to only full-power operation of nuclear power plant (NPP), but some events which were recently occurred could reach severe plant damage state. Thus, various countries around the world have focused their interests on the evaluation for low power/shutdown (LP/S) operation. This report covers the main stream of LP/S PSA methodology, current status of LP/S PSA practices and results, and accident sequence analysis for loss of RHR during mid-loop operation. Therefore this report would be helpful for us to practice LP/S PSA for YGN 5,6 NPP which will be built in the near future. Also the results of accident sequence analysis show that operator's mis-diagnosis and failure of recovery action would initiate core damage during LP/S operation. In summary, overall environmental improvements (equipments, procedures, Tech Spec, etc, ...) and operating support system will be very useful to reduce risk during LP/S operation. (Author) 5 figs., 9 tabs

  20. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  1. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  2. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  3. On-line monitoring for calibration reduction

    International Nuclear Information System (INIS)

    Hoffmann, M.

    2005-09-01

    On-Line Monitoring evaluates instrument channel performance by assessing its consistency with other plant indications. Elimination or reduction of unnecessary field calibrations can reduce associated labour costs, reduce personnel radiation exposure, and reduce the potential for calibration errors. On-line calibration monitoring is an important technique to implement a state-based maintenance approach and reduce unnecessary field calibrations. In this report we will look at how the concept is currently applied in the industry and what the arising needs are as it becomes more commonplace. We will also look at the PEANO System, a tool developed by the Halden Project to perform signal validation and on-line calibration monitoring. Some issues will be identified that are being addressed in the further development of these tools to better serve the future needs of the industry in this area. An outline for how to improve these points and which aspects should be taken into account is described in detail. (Author)

  4. On-line signal trend identification

    International Nuclear Information System (INIS)

    Tambouratzis, T.; Antonopoulos-Domis, M.

    2004-01-01

    An artificial neural network, based on the self-organizing map, is proposed for on-line signal trend identification. Trends are categorized at each incoming signal as steady-state, increasing and decreasing, while they are further classified according to characteristics such signal shape and rate of change. Tests with model-generated signals illustrate the ability of the self-organizing map to accurately and reliably perform on-line trend identification in terms of both detection and classification. The proposed methodology has been found robust to the presence of white noise

  5. On line routing per mobile phone

    DEFF Research Database (Denmark)

    Bieding, Thomas; Görtz, Simon; Klose, Andreas

    2009-01-01

    On-line routing is concerned with building vehicle routes in an ongoing fashion in such a way that customer requests arriving dynamically in time are efficiently and effectively served. An indispensable prerequisite for applying on-line routing methods is mobile communication technology....... Additionally it is of utmost importance that the employed communication system is suitable integrated with the firm’s enterprise application system and business processes. On basis of a case study, we describe in this paper a system that is cheap and easy to implement due to the use of simple mobile phones...

  6. On-line atomic data access

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, D.R. [Oak Ridge National Lab., TN (United States); Nash, J.K. [Lawrence Livermore National Lab., CA (United States)

    1996-04-01

    The need for atomic data is one which continues to expand in a wide variety of applications including fusion energy, astrophysics, laser- produced plasma research, and plasma processing. Modern computer database and communications technology nables this data to be placed on-line and obtained by users of the Internet. Presented here is a summary of the observations and conclusions regarding such on-line atomic data access derived from a forum held at the Tenth APS Topical Conference on Atomic Processes in Plasmas.

  7. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  8. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  9. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  10. On-line monitoring for calibration reduction

    International Nuclear Information System (INIS)

    Hoffmann, Mario; Gran, Frauke Schmitt; Thunem, Harald P-J.

    2004-04-01

    On-Line Monitoring (OLM) of a channel's calibration state evaluates instrument channel performance by assessing its consistency with other plant indications. Industry and experience at several plants has shown this overall approach to be very effective in identifying instrument channels that are exhibiting degrading or inconsistent performance characteristics. The Halden Reactor Project has developed the signal validation system PEANO, which can be used to assist with the tasks of OLM. To further enhance the PEANO System for use as a calibration reduction tool, the following two additional modules have been developed; HRP Prox, which performs pre-processing and statistical analysis of signal data, Batch Monitoring Module (BMM), which is an off-line batch monitoring and reporting suite. The purpose and functionality of the HRP Prox and BMM modules are discussed in this report, as well as the improvements made to the PEANO Server to support these new modules. The Halden Reactor Project has established a Halden On-Line Monitoring User Group (HOLMUG), devoted to the discussion and implementation of on-line monitoring techniques in power plants. It is formed by utilities, vendors, regulatory bodies and research institutes that meet regularly to discuss implementation aspects of on-line monitoring, technical specification changes, cost-benefit analysis and regulatory issues. (Author)

  11. SPIRES I: on-line search guide

    International Nuclear Information System (INIS)

    Addis, L.

    1975-06-01

    SPIRES I is the first generation of the on-line Stanford Public Information Retrieval System. Designed as a prototype system, SPIRES I was later moved to the SLAC computing facility where it has been routinely available to SLAC users in the field of high-energy physics. The scope and use of the SPIRES I system are described in this manual

  12. HOPI: on-line injection optimization program

    International Nuclear Information System (INIS)

    LeMaire, J.L.

    1977-01-01

    A method of matching the beam from the 200 MeV linac to the AGS without the necessity of making emittance measurements is presented. An on-line computer program written on the PDP10 computer performs the matching by modifying independently the horizontal and vertical emittance. Experimental results show success with this method, which can be applied to any matching section

  13. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  14. A development of maintenance educational support method by using navigation method Pt. 4. A development of a paper prototype system for the RHR pump maintenance educational support

    International Nuclear Information System (INIS)

    Yoshino, Kenji; Hirotsu, Yuko; Fujimoto, Junzo; Tsumura, Joji

    2004-01-01

    This research establishes the new maintenance educational support prototype system for aiming at organization as Trinity (a supervision person, a construction person in charge, a work person in charge, and group length), and improvement in human reliability for the purpose of aiming at before hand prevention of a systematic error, human error, and the accident and a trouble. (1) This maintenance educational support: system made educational start to the educational end the structure navigated exactly, carrying out the exchange with a data base along with the road map which divided action (act) of the work process and worker concerning RHR pump maintenance work for a participant etc. by class. (2) First, according to a participant's attribute, this educational support system performs educational support (a question and reply) concerning an organization and organization, and judges knowledge and a nature type based on the result. Furthermore, educational support (a question and reply) concerning maintenance work is performed, and a knowledge level is judged based on the result. And it considered as the structure which performs the comprehensive judgment about an organization, organization, and the educational effect concerning maintenance work at the last. (3) The judgment pattern of the knowledge and nature after the educational support end concerning a participant's organization and organization was made into four kinds which consist of (1) organization priority and duty A type, (2) organization priority and an adaptable type, (3) individual priority and an adaptable type, and (4) individual priority and a duty priority type. (4) The judgment pattern of the knowledge level after the educational support end concerning maintenance work of a participant was used as three kinds of evaluation axes, schedule management, apparatus reliability capability, and human-being reliability. (author)

  15. Educational On-Line Gaming Propensity

    DEFF Research Database (Denmark)

    Sudzina, Frantisek; Razmerita, Liana; Kirchner, Kathrin

    2014-01-01

    Educational on-line games are promising for new generations of students who are grown up digital. Th e new generations of students are technology savvy and spend lots of time on the web and on social networks. Based on an exploratory study, this article investigates the factors that infl uence...... students’ willingness to participate in serious games for teaching/learning. Th is study investigates the relationship between students’ behavior on Facebook, Facebook games, and their attitude toward educational on-line games. Th e results of the study reveal that the early adopters of educational games...... are likely to be students, who are young, have only a few Facebook connections, who currently play Facebook game(s). Furthermore, the study emphasizes that there may be differences between students coming from various countries....

  16. Trends in on-line data processing

    International Nuclear Information System (INIS)

    Masetti, M.

    1981-01-01

    The developement of integrated circuits has been characterized by an exponential growth of gates on a single chip that will still continue in the coming years. In parallel the price per bit is dropping down with more or less the same law. As a consequence of this a few statements can be made: The present 16-bit minicomputer in a small configuration is going to be substituted by a 16-bit microcomputer, and the 16-bit microcomputer in a powerful configuration by a 32-bit midi having also a virtual memory facility. Fully programmable or microcoded powerful devices like the LASS hardware processor or MICE, will allow an efficient on-line filter. Higher computing speed can be achieved by a multiprocessor configuration which can be insensitive to hardware failures. Therefore we are moving towards an integrated on-line computing system with much higher computing power than now and the present distinction between on-line and off-line will no longer be so sharp. As more processing can be performed on-line, fast high quality feed-back can be provided for the experiment. In the years to come the trend towards more processing power, at a lower price, and assembled in the same hardware volume will continue for at least five years; at the same time the future large high-energy physics experiments at LEP will be carried out within a wide international collaboration. In this environment methods must be found for a large fraction of the work to be distributed amongst the collaborators. To accomplish this aim it is necessary to introduce common standard practices concerning both hardware and software, in such a way that the seperate parts, developed by the collaborators, will be plug-compatible. (orig.)

  17. New Trends in on-line Marketing

    OpenAIRE

    Palkovič, Lukáš

    2011-01-01

    This bachelor thesis deals with new trend of internet marketing, it focuses especially on viral marketing. The theoretical part charasterizes the process of viral campaigns, furthermore deals with the components and aspects of on-line environment. Another separated chapter presents social networks, their place in viral marketing and at last but not least the viral video making process. The practical part contains different analyses of specific viral campaigns. The next and equally the last pa...

  18. On-Line Maintenance Methodology Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyo Won; Kim, Jae Ho; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    Most of domestic maintenance activities for nuclear power plants are performed while overhaul. Therefore, On-Line Maintenance (OLM) is one of the proper risks informed application techniques for diffusing maintenance burden during overhaul with safety of the plant is secured. The NUMARC 93-01 (Rev.3) presents the OLM state of the art and it provides methodology. This study adopts NUMARC 93-01 (Rev.3) and present OLM. The reference component is Emergency Diesel Generator (EDG) of Ulchin 3, 4

  19. Demonstration of an automated on-line surveillance system at a commercial nuclear power plant

    International Nuclear Information System (INIS)

    Smith, C.M.; Sweeney, F.J.

    1983-01-01

    As a first step in demonstrating the practicality of performing continuous on-line surveillance of the performance of nuclear steam supply systems using noise related techniques, Oak Ridge National Laboratory is operating a computerized noise signal data acquisition and processing system at the Sequoyah Unit 1 Nuclear Plant, an 1148 MWe four-loop Westinghouse pressurized water reactor (PWR) located near Chattanooga, Tennessee. The principal objective is to establish, with a degree of continuity and completeness not previously achieved, the long-term characteristics of signals from neutron detectors and process sensors in order to evaluate the feasibility of detecting and diagnosing anomalous reactor conditions by means of changes in these signals. The system is designed to automatically screen the gathered data, using a number of descriptors derived from the power spectra of the monitored signals, and thereby select for the noise analyst's perusal only those data which differ statistically from norms which the system has previously established

  20. Experiences with 'on-line' diagnostic instrumentation in nuclear power plants

    International Nuclear Information System (INIS)

    Gopal, R.; Ciaramitaro, W.; Smith, J.R.

    1981-01-01

    Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve Nuclear Steam Supply System (NSSS) availability and maintainability through the detection of malfunctions at their inception. These systems include: 1) Acoustic leak monitoring for detection and location of leaks in the primary system pressure boundary and other piping systems in PWR's; 2) Metal impact monitoring for detection of loose debris in the reactor vessel and steam generators; 3) Nuclear noise monitoring for monitoring core barrel vibration. Summarized in this paper are some of the features of the systems and inplant experience. (author)

  1. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  2. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  3. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  4. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  5. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  6. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  7. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  8. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  9. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  10. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  11. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    International Nuclear Information System (INIS)

    Nabeshima Kunihiko; Suzuki Katsuo; Nose, Shoichi; Kudo, Kazuhiko

    1996-01-01

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs

  12. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kunihiko, Nabeshima; Katsuo, Suzuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Nose, Shoichi; Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-12-31

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs.

  13. Chained computations using an unsteady 3D approach for the determination of thermal fatigue in a T-junction of a PWR nuclear plant

    International Nuclear Information System (INIS)

    Pasutto, Thomas; Peniguel, Christophe; Sakiz, Marc

    2006-01-01

    Thermal fatigue of the coolant circuits of PWR plants is a major issue for nuclear safety. The problem is especially accute in mixing zones, like T-junctions, where large differences in water temperature between the two inlets and high levels of turbulence can lead to large temperature fluctuations at the wall. Until recently, studies on the matter had been tackled at EDF using steady methods: the fluid flow was solved with a CFD code using an averaged turbulence model, which led to the knowledge of the mean temperature and temperature variance at each point of the wall. But, being based on averaged quantities, this method could not reproduce the unsteady and 3D effects of the problem, like phase lag in temperature oscillations between two points, which can generate important stresses. Benefiting from advances in computer power and turbulence modeling, a new methodology is now applied, that allows to take these effects into account. The CFD tool Code S aturne, developed at EDF, is used to solve the fluid flow using an unsteady L.E.S. approach. It is coupled with the thermal code Syrthes, which propagates the temperature fluctuations into the wall thickness. The instantaneous temperature field inside the wall can then be extracted and used for structure mechanics computations (mainly with EDF thermomechanics tool Code A ster). The purpose of this paper is to present the application of this methodology to the simulation of a straight T-junction mock-up, similar to the Residual Heat Remover (RHR) junction found in N4 type PWR nuclear plants, and designed to study thermal striping and cracks propagation. The results are generally in good agreement with the measurements; yet, in certain areas of the flow, progress is still needed in L.E.S. modelling and in the treatment of instantaneous heat transfer at the wall

  14. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  15. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Amey, M.D.H.; Brown, G.R.

    1987-01-01

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  16. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Nordmann, F.; Odar, S.; Rochester, D.

    2012-01-01

    Secondary side degradation of steam generators (SG) tubing with Alloy 600 MA and flow accelerated corrosion (FAC) of carbon steel have been for a long time important issues for the secondary system of PWR and VVER. With the beneficial evolution of the design (for instance the replacement of Alloy 600 SG tubing), the most important issues are progressively moving to a larger variety of risks associated to potential inadequate chemistries. The best remedies for mitigating the new concerns are: -) selecting a steam water treatment able to minimize the quantity of corrosion products transported to the steam generator, -) mitigating the risk of flow induced vibration by a proper control of deposits in sensitive areas, -) minimizing the risk of concentration of impurities in local areas where they may induce corrosion. The paper also explains: -) the benefit of eliminating or by pass of condensate polishers, -) the absence of need for expensive lead investigation, if no specific pollution occurred, -) the absence of need for very low oxygen in the condensate water, and -) the necessary and optimum number of on-line monitors

  17. On-line reconstruction of in-core power distribution by harmonics expansion method

    International Nuclear Information System (INIS)

    Wang Changhui; Wu Hongchun; Cao Liangzhi; Yang Ping

    2011-01-01

    Highlights: → A harmonics expansion method for the on-line in-core power reconstruction is proposed. → A harmonics data library is pre-generated off-line and a code named COMS is developed. → Numerical results show that the maximum relative error of the reconstruction is less than 5.5%. → This method has a high computational speed compared to traditional methods. - Abstract: Fixed in-core detectors are most suitable in real-time response to in-core power distributions in pressurized water reactors (PWRs). In this paper, a harmonics expansion method is used to reconstruct the in-core power distribution of a PWR on-line. In this method, the in-core power distribution is expanded by the harmonics of one reference case. The expansion coefficients are calculated using signals provided by fixed in-core detectors. To conserve computing time and improve reconstruction precision, a harmonics data library containing the harmonics of different reference cases is constructed. Upon reconstruction of the in-core power distribution on-line, the two closest reference cases are searched from the harmonics data library to produce expanded harmonics by interpolation. The Unit 1 reactor of DayaBay Nuclear Power Plant (DayaBay NPP) in China is considered for verification. The maximum relative error between the measurement and reconstruction results is less than 5.5%, and the computing time is about 0.53 s for a single reconstruction, indicating that this method is suitable for the on-line monitoring of PWRs.

  18. Field experience with advanced methods of on-line monitoring of water chemistry and corrosion degradation in nuclear power stations

    International Nuclear Information System (INIS)

    Stellwag, B.; Aaltonen, P.; Hickling, J.

    1997-01-01

    Advanced methods for on-line, in-situ water chemistry and corrosion monitoring in nuclear power stations have been developed during the past decade. The terms ''on-line'' and ''in-situ'' characterize approaches involving continuous measurement of relevant parameters in high temperature water, preferably directly in the systems and components and not in removed samples at room temperature. This paper describes the field experience to-date with such methods in terms of three examples: (1) On-line chemistry monitoring of the primary coolant during shutdown of a Type WWER-440 PWR. (2) Redox and corrosion potential measurements in final feedwater preheaters and steam generators of two large KWU PWRs over several cycles of plant operation. (3) Real-time, in-situ corrosion surveillance inside the calundia vault of a CANDU reactor. The way in which water chemistry sensors and corrosion monitoring sensors complement each other is outlined: on-line, in-situ measurement of pH, conductivity and redox potential gives information about the possible corrosivity of the environment. Electrochemical noise techniques display signals of corrosion activity under the actual environmental conditions. A common experience gained from separate use of these different types of sensors has been that new and additional information about plants and their actual process conditions is obtained. Moreover, they reveal the intimate relationship between the operational situation and its consequences for the quality of the working fluid and the corrosion behaviour of the plant materials. On this basis, the efficiency of the existing chemistry sampling and control system can be checked and corrosion degradation can be minimized. Furthermore, activity buildup in the primary circuit can be studied. Further significant advantages can be expected from an integration of these various types of sensors into a common water chemistry and corrosion surveillance system. For confirmation, a complete set of sensors

  19. On-line neutron activation analyzers

    International Nuclear Information System (INIS)

    Flahaut, V.; Colmon, A.

    1999-01-01

    A neutronic analyser has been designed to determine the composition of the flow of raw materials entering a cement factory on the conveyor belt. This new system gives a reliable analysis of the whole cargo that outdates the sampling or the usual surface analysis based on fluorescence spectrometry. The accuracy is about 1%.The neutrons interact with the materials on an average depth of 25 cm and are absorbed by nuclei, these nuclei produce photons whose energy is characteristic of the chemical element itself. The composition can be deduced by measuring the number of photons emitted and their energy. The analysis is made on-line and can concern the search for about 10 compounds. In the case of cement the list of compounds is: SiO 2 , CaO, Al 2 O 3 , Fe 2 O 3 , MgO, Na 2 O, TiO 2 , S, Mn 2 O 3 , K 2 O, and H 2 O. The neutron generator involves a deuterium ion source whose deuterium ions are accelerated by means of an electrical field and impinge on a tritiated target, the nuclear reactions between deuterium and tritium produce 14 MeV neutrons. This neutron analysing technique can be adapted to any need of on-line composition determination. (A.C.)

  20. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  1. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  2. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  3. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  4. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  5. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  6. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  7. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  8. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  9. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  10. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  11. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  12. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  13. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  14. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  15. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  16. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  17. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  18. On line nuclear orientation: opportunity and challenge

    International Nuclear Information System (INIS)

    Stone, N.J.

    1985-01-01

    The development of the on-line nuclear orientation (OLNO) technique is reviewed. The present potential of the technique is discussed in the light of the attainable temperatures, the use of ion implantation and the required isotope flux. Limitations associated with spin-lattice relaxation are considered in some detail and a survey of accessible nuclei is presented. An outline comparison is given between OLNO and other methods for producing orientation of nuclei, for measuring nuclear spins and static moments and for the study of level structure and transition probabilities. The conclusion is drawn that the method in its present form has extensive potential over a wide range of nuclei. Future prospects for in-beam polarisation giving access to nuclei of shorter half lives are referred to briefly. (Auth.)

  19. An on-line diagnostic expert system

    International Nuclear Information System (INIS)

    Felkel, L.

    1987-01-01

    As experience with on-line information systems, experts systems and artificial intelligence tools grows, the authors retreat from the first euphoria that AI could help them solve the problem they were unable to solve with conventional programming. The major effort of the development time goes into building the knowledge-base. There is no such thing as a generic knowledge-base for nuclear power plants as there is, for example, for the diagnosis of a Boeing 747 aircraft. AI-methods, tools and hardware are still in a state which does not optimally lend itself to real-time application. The ability of developing prototype systems to investigate variants otherwise too costly to justify is one advantage that the authors gladly accept. Last, but no least the tools provide a flexible and adaptable user interface (desktop window systems) etc. The development of such tools in a project would be prohibitive and room for experimentation would be limited

  20. On-Line Algorithms and Reverse Mathematics

    Science.gov (United States)

    Harris, Seth

    In this thesis, we classify the reverse-mathematical strength of sequential problems. If we are given a problem P of the form ∀X(alpha(X) → ∃Zbeta(X,Z)) then the corresponding sequential problem, SeqP, asserts the existence of infinitely many solutions to P: ∀X(∀nalpha(Xn) → ∃Z∀nbeta(X n,Zn)). P is typically provable in RCA0 if all objects involved are finite. SeqP, however, is only guaranteed to be provable in ACA0. In this thesis we exactly characterize which sequential problems are equivalent to RCA0, WKL0, or ACA0.. We say that a problem P is solvable by an on-line algorithm if P can be solved according to a two-player game, played by Alice and Bob, in which Bob has a winning strategy. Bob wins the game if Alice's sequence of plays 〈a0, ..., ak〉 and Bob's sequence of responses 〈 b0, ..., bk〉 constitute a solution to P. Formally, an on-line algorithm A is a function that inputs an admissible sequence of plays 〈a 0, b0, ..., aj〉 and outputs a new play bj for Bob. (This differs from the typical definition of "algorithm", though quite often a concrete set of instructions can be easily deduced from A.). We show that SeqP is provable in RCA0 precisely when P is solvable by an on-line algorithm. Schmerl proved this result specifically for the graph coloring problem; we generalize Schmerl's result to any problem that is on-line solvable. To prove our separation, we introduce a principle called Predictk(r) that is equivalent to -WKL0 for standard k, r.. We show that WKL0 is sufficient to prove SeqP precisely when P has a solvable closed kernel. This means that a solution exists, and each initial segment of this solution is a solution to the corresponding initial segment of the problem. (Certain bounding conditions are necessary as well.) If no such solution exists, then SeqP is equivalent to ACA0 over RCA 0 + ISigma02; RCA0 alone suffices if only sequences of standard length are considered. We use different techniques from Schmerl to prove

  1. Software for on-line experiments

    International Nuclear Information System (INIS)

    Ivanchenko, I.M.

    1981-01-01

    A review of nowadays development state of software of on-line electron experiments is presented. The principles of organization of real time systems on second generation computer base are considered. The following methods for projections search are considered: combinator methods, global methods, methods of tracking, methods of a supporting band. The following methods for determining parameter estimates based on the Lorentz equation are analysed: analytical simulation of trajectories, determination of parameters by the iterative method using the technique of calculation of recycled integrals, multidimensional statistical analysis. For the purpose of successful usage and development of software the technique of selfdocumented programs is created and the computer is applied for preparing, revising and circulation of external descriptions which as program complexes are constructed according to the hierarchical principle [ru

  2. On line protection systems for induction motors

    International Nuclear Information System (INIS)

    Colak, I.; Celik, H.; Sefa, I.; Demirbas, S.

    2005-01-01

    Protection of induction motors is very important since they are widely used in industry for many applications due to their high robustness, reliability, low cost and maintenance, high efficiency and long service life. So, protecting these motors is crucial for operations. This paper presents a combined protection approach for induction motors. To achieve this, the electrical values of the induction motor were measured with sensitivity ±1% through a data acquisition card and processed with software developed in Visual C++. An on line protection system for induction motors was achieved easily and effectively. The experimental results have shown that the induction motor was protected against the possible problems faced during the operation. The software developed for this protection provides flexible and reliable media for operators and their motors. It is expected that the motor protection achieved in this study might be faster than the classical techniques and also may be applied to larger motors easily after small modifications of the software

  3. Aprender a innovar: una experiencia on line

    Directory of Open Access Journals (Sweden)

    Joaquín MORENO MARCHAL

    2014-11-01

    Full Text Available La creatividad y la innovación se han convertido en recursos clave en la denominada sociedad del conocimiento, que bien podría ser también llamada sociedad de la innovación. Pero innovar es una actividad compleja, que integra la aplicación de múltiples capacidades, el pensamiento divergente y convergente, la gestión de equipos humanos, la comunicación. Ahora bien, a innovar se puede, y se debe, aprender. Aprender a innovar es un reto y también una obligación para el conjunto del sistema educativo en todos sus niveles. Partiendo de estas consideraciones este trabajo expone una experiencia de aprendizaje de la creatividad y de la innovación a través de un curso totalmente on line basado en la plataforma MOODLE, en el marco del Programa de Formación Permanente de la Universidad de Cádiz. Se presenta un modelo del proceso de innovación, denominado CREALAB, de elaboración propia. Este modelo se ha utilizado como base del proceso de aprendizaje de la creatividad y de la innovación y en el diseño del curso, está organizado en torno a actividades y tiene un carácter iterativo y realimentado. Se presentan además el conjunto del diseño metodológico y los resultados obtenidos en las dos ediciones celebradas hasta el momento. El diseño del curso totalmente on line y los resultados alcanzados permiten estimar un alto potencial de aplicación, tanto a nivel personal como a nivel organizacional.

  4. On-line diagnostic techniques for air-operated control valves based on time series analysis

    International Nuclear Information System (INIS)

    Ito, Kenji; Matsuoka, Yoshinori; Minamikawa, Shigeru; Komatsu, Yasuki; Satoh, Takeshi.

    1996-01-01

    The objective of this research is to study the feasibility of applying on-line diagnostic techniques based on time series analysis to air-operated control valves - numerous valves of the type which are used in PWR plants. Generally the techniques can detect anomalies by failures in the initial stages for which detection is difficult by conventional surveillance of process parameters measured directly. However, the effectiveness of these techniques depends on the system being diagnosed. The difficulties in applying diagnostic techniques to air-operated control valves seem to come from the reduced sensitivity of their response as compared with hydraulic control systems, as well as the need to identify anomalies in low level signals that fluctuate only slightly but continuously. In this research, simulation tests were performed by setting various kinds of failure modes for a test valve with the same specifications as of a valve actually used in the plants. Actual control signals recorded from an operating plant were then used as input signals for simulation. The results of the tests confirmed the feasibility of applying on-line diagnostic techniques based on time series analysis to air-operated control valves. (author)

  5. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  6. Issues and remedies for secondary system of PWR/VVER

    International Nuclear Information System (INIS)

    Nordmann, Francis; Odar, Suat; Rochester, Dewey

    2012-09-01

    ); - Applying preventive remedies such as soft SG cleanings and dispersant addition for avoiding expensive curative cleanings; - Optimizing operating costs and release of effluents into the environment. This paper also explains the past issues that do not deserve any more of the same drastic efforts according to the design evolution. For example, the paper explains the: - Benefit of elimination or by pass of condensate polishers; - Absence of need for expensive lead investigation, if no specific pollution occurred; - Absence of need for very low oxygen in the condensate water; - Necessary and optimum number of on-line monitors. (authors)

  7. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  8. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  9. Feasibility study of applying a multi-channel analysis model to on-line core monitoring system

    International Nuclear Information System (INIS)

    In, W. K.; Yoo, Y. J.; Hwang, D. H.; Jun, T. H.

    1998-01-01

    A feasibility study was performed to evaluate the effect of implementing a multi-channel analysis model in on-line core monitoring system. A simplified thermal-hydraulic model has been used in the on-line core monitoring system of digital PWR. The design procedure, core thermal margin and computation time were investigated in case of replacing the simplified model with the multi-channel analysis model. For the given ranges of limiting conditions for operation in Yonggwang Unit 3 Cycle 1, the minimum DNBR of the simplified thermal-hydraulic code CETOP-D was compared to that of the multi-channel analysis code MATRA. A CETOP-D tuning is additionally required to ensure the accurate and conservative DNBR calculation but the MATRA tuning is not necessary. MATRA appeared to increase the DNBR overpower margin from 2.5% to 6% over the CETOP-D margin. MATRA took approximately 1 second to compute DNBR on the HP9000 workstation system, which is longer than the DNBR computation time of CETOP-D. It is, however, fast enough to perform the on-line monitoring of DNBR. It can be therefore concluded that the application of the multi-channel analysis model MATRA in the on-line core monitoring system is feasible

  10. On-line maintenance at Cofrentes NPP

    International Nuclear Information System (INIS)

    Roldan Vilches, J.; Moreno Matarranz, M. A.; Hermana Mendioroz, I.

    1998-01-01

    Cofrentes NPP has begun in 1997 activities related to At Power Preventive Maintenance over trains or systems which lead to a voluntary entry in a Limitative Condition of Operation (LCO) of the Technical Specifications. From others benefits, this program ha improved the risk management and the staff's knowledge over the functions and safety implications of the different systems, the better exploit of the resources, the co-ordination of the different organisations involved (Maintenance an Operation) and the reductions of works during shutdowns. Previous to each work, a feasibility study analyzes qualitative and quantitative (PSA), using the Risk Monitor, the implications on safety of all the tasks, assuring that the global safety of the Plant is always maintained. Tech. Spec. are analyzed in detail and also are analyzed situations of simultaneous unavailabilities of systems which could lead to a high risk situation. Two different risk controls are defined (punctual and accumulated) to assure that high risk situations will not be given. Finally, historical risk profile is analyzed to assure that the accumulated risk increase is not significant. Risk Monitor helps staff in the schedule and follow-up of the activities of On-Line Maintenance. Each one of the tasks are deeply planned and harshly analyzed and are carried out by high qualified workers. By the moment, this program is running with fully satisfaction on the Plant. (Author)

  11. SOL: INNOVACIÓN ON-LINE

    Directory of Open Access Journals (Sweden)

    Rubén Faúndez

    2007-11-01

    Full Text Available Las aplicaciones de simulación tienden a ser cada vez más cercanas a usuarios e industrias. Sin embargo, muchas de ellas no poseen ni la capacidad ni el conocimiento como para desarrollar internamente sus modelos de simulación. Por este motivo, y como una forma de apoyar la toma de decisiones basándose en modelos de simulación, se presenta la plataforma SOL (Simulación On Line. La metodología completa de trabajo, así como la interacción entre SOL, Empresa y Asesor, son presentadas. Su base de datos, los niveles de usuarios, sus funcionalidades, y la creación automatizada de información grafica y visual, también son explicadas. En el caso de aplicación, el uso de SOL para apoyar la toma de decisiones en una operación de movimiento de material, permite a los tomadores de decisión acceder a análisis robustos basados en información extraída de los modelos de simulación. SOL, al almacenar información, funcionar vía web, generar análisis automatizados y crear visualizaciones, permite cumplir con las expectativas de los usuarios respecto a una solución integral en simulación.

  12. Simplified automatic on-line document searching

    International Nuclear Information System (INIS)

    Ebinuma, Yukio

    1983-01-01

    The author proposed searching method for users who need not-comprehensive retrieval. That is to provide flexible number of related documents for the users automatically. A group of technical terms are used as search terms to express an inquiry. Logical sums of the terms in the ascending order of frequency of the usage are prepared sequentially and automatically, and then the search formulas, qsub(m) and qsub(m-1) which meet certain threshold values are selected automatically also. Users justify precision of the search output up to 20 items retrieved by the formula qsub(m). If a user wishes more than 30% of recall ratio, the serach result should be output by qsub(m), and if he wishes less than 30% of it, it should be output by qsub(m-1). The search by this method using one year volume of INIS Database (76,600 items) and five inquiries resulted in 32% of recall ratio and 36% of precision ratio on the average in the case of qsub(m). The connecting time of a terminal was within 15 minutes per an inquiry. It showed more efficiency than that of an inexperienced searcher. The method can be applied to on-line searching system for database in which natural language only or natural language and controlled vocabulary are used. (author)

  13. On-Line Impact Load Identification

    Directory of Open Access Journals (Sweden)

    Krzysztof Sekuła

    2013-01-01

    Full Text Available The so-called Adaptive Impact Absorption (AIA is a research area of safety engineering devoted to problems of shock absorption in various unpredictable scenarios of collisions. It makes use of smart technologies (systems equipped with sensors, controllable dissipaters and specialised tools for signal processing. Examples of engineering applications for AIA systems are protective road barriers, automotive bumpers or adaptive landing gears. One of the most challenging problems for AIA systems is on-line identification of impact loads, which is crucial for introducing the optimum real-time strategy of adaptive impact absorption. This paper presents the concept of an impactometer and develops the methodology able to perform real-time impact load identification. Considered dynamic excitation is generated by a mass M1 impacting with initial velocity V0. An analytical formulation of the problem, supported with numerical simulations and experimental verifications is presented. Two identification algorithms based on measured response of the impacted structure are proposed and discussed. Finally, a concept of the AIA device utilizing the idea of impactometer is briefly presented.

  14. Application of on-line analytical processing technique in accelerator

    International Nuclear Information System (INIS)

    Xie Dong; Li Weimin; He Duohui; Liu Gongfa; Xuan Ke

    2005-01-01

    A method of application of the on-line analytical processing technique in accelerator is described, which includes data pre-processing, the process of constructing of data warehouse and on-line analytical processing. (authors)

  15. Simultaneous Determination of Palladium and Platinum by On-line ...

    African Journals Online (AJOL)

    NJD

    using high performance liquid chromatography equipped with an on-line enrichment technique. Prior to ... The on-line enrichment system (Waters Corporation, USA) that was ... Using an appropriate volume (industrial plant effluents 20 mL,.

  16. Influence of visualization on consumption during on-line shopping

    OpenAIRE

    Hictaler, Urška

    2013-01-01

    This diploma work studies the influence of visualization on consumption during on-line shopping. The first part of the thesis starts with key areas of visualization, consumption and on-line shopping. Visualization, areas of use, human perception and ways of product presentation in on-line shops are defined discussed first. Next, consumption, consumers and factors that influence their decisions and satisfaction are defined. The last topic in the first part of the thesis discusses on-line shopp...

  17. A Distributed System for Learning Programming On-Line

    Science.gov (United States)

    Verdu, Elena; Regueras, Luisa M.; Verdu, Maria J.; Leal, Jose P.; de Castro, Juan P.; Queiros, Ricardo

    2012-01-01

    Several Web-based on-line judges or on-line programming trainers have been developed in order to allow students to train their programming skills. However, their pedagogical functionalities in the learning of programming have not been clearly defined. EduJudge is a project which aims to integrate the "UVA On-line Judge", an existing…

  18. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  19. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  20. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  1. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  2. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  3. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  4. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  5. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  6. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  7. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  8. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  9. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  10. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  11. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  12. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  13. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  14. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  15. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  16. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  17. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  18. Radiation detectors for the control of PWR nuclear boilers

    International Nuclear Information System (INIS)

    Duchene, J.

    1977-01-01

    The neutronic control in French PWR is effected by: 2 channels of measurement of intermediate power using γ'-compensated boron-coated ionization chambers 4 channels of measurement of high power with 'long' boron chambers also used in axial off-set measurement. A movable in-core measuring system is used for the fuel management and the power distribution monitoring. The instrumentation of start-up and intermediate power is conventional; the chambers of the axial off-set measurement and the in-core system are special for this type of power plant, they are discussed in details. The essential properties of the various types of detector, their major advantages or drawbacks, their comparative adaptation to the functions to be performed in the plant are summarized in a table. The 'long chambers' (on use in Fessenheim I and II, and soon in Bugey II) are boron coated current ionization chambers, without γ compensation, intended for power measurement. In-core measurements first involved activation methods - movable wires giving flux profiles, -or activable nuts (the Aeroball System at Trino Vercellese, Chooz...). In on-line neutron detectors, used at fixed positions, the electric signal is generated from: ionization the gas filling fission ionization chambers and γ ionization chambers; direct collection of the charged particles emitted from the convertor element in self-powered neutron detectors (rhodium, silver or vanadium) or self-powered γ detectors (cobalt); or thermoelectric effect in neutron and γ thermometers. The in-core measurement unit developped by Framatome is a movable miniaturized fission chamber system (at Tihange), every French exported power plant being now equipped with it [fr

  19. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  20. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  1. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  2. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  3. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  4. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  5. Robust on-line monitoring of biogas processes; Robusta maettekniker on-line foer optimerad biogasproduktion

    Energy Technology Data Exchange (ETDEWEB)

    Nordberg, Aake; Hansson, Mikael; Kanerot, Mija; Krozer, Anatol; Loefving, Bjoern; Sahlin, Eskil

    2010-03-15

    Although demand for biomethane in Sweden is higher than ever, many Swedish codigestion plants are presently operated below their designed capacity. Efforts must be taken to increase the loading rate and guarantee stable operation and high availability of the plants. There are currently no commercial systems for on-line monitoring, and due to the characteristics of the material, including corrosion and tearing, robust applications have to be developed. The objective of this project was to identify and study different monitoring technologies with potential for on-line monitoring of both substrate mixtures and anaerobic digester content. Based on the prerequisites and demands at Boraas Energi och Miljoe AB's (BEMAB, the municipal energy and waste utility in the city of Boraas, Sweden) biogas plant, the extent of the problems, measurement variables and possible ways of managing these issues have been identified and prioritized. The substrate mixtures in question have a high viscosity and are inhomogeneous with variation in composition, which calls for further homogenization, dilution and filtration to achieve high precision in the necessary analyses. Studies of using different mixers and mills showed that the particle size (800 mum) needed for on-line COD measurement could not be achieved. The problem of homogenization can be avoided if indirect measurement methods are used. Laboratory tests with NIR (near-infra red spectroscopy) showed that VS can be predicted (R2=0,78) in the interval of 2-9% VS. Furthermore, impedance can give a measurement of soluble components. However, impedance is not sensitive enough to give a good measurement of total TS. Microwave technology was installed at the production plant and showed a faster response to changes in TS than the existing TS-sensor. However, due to technical problems, the evaluation only could be done during a limited period of ten days. BEMAB will continue the measurements and evaluation of the instrument. The

  6. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  7. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  8. On-line fast flux measurements in the BR2 reactor

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2009-01-01

    Since 2001, CEA-Cadarache and the Belgian Nuclear Research Centre SCK-CEN are collaborating on the development and in-pile qualification of subminiature fission chambers (diameter of 1.5 mm). Initially, efforts concentrated on fission chambers for the in-pile measurement of thermal fluxes (with 235 U as fissile material). Meanwhile successful long-term tests of the prototypes have been performed in various environments: in low temperature (40-100 degress Celsius) BR2 pool water (up to a thermal neutron fluence of 3 1 0 21 n/cm 2 ) and in the CALLISTO PWR loop (300 degrees Celsius, 155 bars). The long-term qualification of derived industrial detectors (Photonis CFUZ53) in CALLISTO is still ongoing. However, for various types of irradiations in research reactors, the knowledge of the evolution of the fast neutron flux is even of more interest than the thermal flux data. Therefore the collaboration program was extended to the development and the in-pile qualification of subminiature or miniature fission chambers (with 3 mm diameter) for fast neutron detection, for which 242 Pu was selected as the optimal fissile material. In order to achieve the on-line in-pile measurement of fast neutron flux, the fission chambers will be operated in the Campbelling mode (based on the mean square fluctuation of the detector current). In this mode the gamma induced contribution to the signal can be efficiently suppressed. Moreover, a data processing software will take into account the evolution of the fissile deposit in order to assess on-line the fast flux sensitivity and to correct for the low energy neutron contributions. The final objective is to qualify a Fast Neutron Detector System (FNDS) able to provide on-line data for local fast neutron fluxes in Material Testing Reactors. The on-line measurement of the fast neutron flux would contribute significantly to the characterization of the irradiation conditions during test experiments with materials and innovative fuel elements

  9. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  10. DOE-EPRI On-Line Monitoring Implementation Guidelines

    International Nuclear Information System (INIS)

    E. Davis, R. Bickford

    2003-01-01

    Industry and EPRI experience at several plants has shown on-line monitoring to be very effective in identifying out-of-calibration instrument channels or indications of equipment-degradation problems. The EPRI implementation project for on-line monitoring has demonstrated the feasibility of on-line monitoring at several participating nuclear plants. The results have been very encouraging, and substantial progress is anticipated in the coming years

  11. A universal PWR spectral history correction

    International Nuclear Information System (INIS)

    Hutt, P.K.; Nunn, D.L.

    1989-01-01

    The accuracy of a form of universal correction for the difference between depletion conditions assumed in PWR assembly lattice calculations and those experienced in a reactor burn-up is investigated. The correction is based on lattice calculations in which only one such depletion history difference, depletion at two different water densities, is explicitly represented by lattice calculations. The assumption is made that other historical effects bear the same relationship to an appropriate time-average of the two-group neutron flux spectrum. The correction is shown to be accurate for the most important historical effects, depletion with burnable absorbers inserted, control rods inserted or at a different soluble boron level, in addition to density itself. The correction is less accurate for representing depletion at a different fuel or coolant temperature but even in these cases gives an improvement over no correction. In addition it is argued that these historic temperature effects are likely to be of minor importance. (author)

  12. Evaluation model for PWR irradiated fuel

    International Nuclear Information System (INIS)

    Gomes, I.C.

    1983-01-01

    The individual economic value of the plutonium isotopes for the recycle of the PWR reactor is investigated, assuming the existence of an market for this element. Two distinct market situations for the stages of the fuel cycle are analysed: one for the 1972 costs and the other for costs of 1982. Comparisons are made for each of the two market situations concerning enrichment of the U-235 in the uranium fuel that gives the minimum cost in the fuel cycle. The method adopted to establish the individual value of the plutonium isotopes consists on the economical analyses of the plutonium fuel cycle for four different isotopes mixtures refering to the uranium fuel cycle. (Author) [pt

  13. Chemical cleaning of nuclear (PWR) steam generators

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Mundis, J.A.

    1982-01-01

    This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

  14. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  15. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  16. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  17. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    Cao, L.; Wu, H.; Zheng, Y.

    2012-01-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  18. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  19. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  20. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  1. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  2. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

    In response to the increased power of the advanced PWR, it is necessary to develop a steam generator (SG) which has a large capacity with high performance and high reliability as well as being economical to produce. In this paper, the development of the design of a new SG for the advanced PWRs is described and compared with the design of a conventional SG. Moreover, an outline of a seismic verification test for the U-bend tube bundle which includes advanced anti-vibration bars (AVB) which are very important is described. As a result, it was verified that the bundle has sufficient strength and a relatively high attenuation to seismic loads. These results will be reflected in the detailed design of advanced AVBs. (author)

  3. Radiation embrittlement of PWR vessel supports

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

    1989-01-01

    Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

  4. Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

    International Nuclear Information System (INIS)

    Chiu, C.

    1981-01-01

    Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nuclear Boiling Ratio (DNBR) which ultimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithm. First, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for this channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum. Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties. Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code. (orig.)

  5. Changes in 900 MW PWR alarm processing policy

    Energy Technology Data Exchange (ETDEWEB)

    Pont, M [Electricite de France, Generation and Transmission, Nuclear Power Plant Operations, Paris (France)

    1997-09-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs.

  6. Changes in 900 MW PWR alarm processing policy

    International Nuclear Information System (INIS)

    Pont, M.

    1997-01-01

    Following a brief description of the current 900 MW PWR alarm processing system, this document presents the feasibility study carried out within the scope of the Instrumentation and Control Refurbishment project (R2C). (author). 4 figs, tabs

  7. Deboration in nuclear stations of the PWR type

    International Nuclear Information System (INIS)

    1978-01-01

    Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

  8. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  9. Dose rate evaluation after accident in a PWR

    International Nuclear Information System (INIS)

    Cladel, C.; Duchemin, B.; Le Dieu de Ville, A.; Nimal, B.; Nimal, J.C.; Evrard, J.M.

    1983-05-01

    A calculation scheme for the gamma radiation dose rate after accident in a PWR is presented. These studies use a fine description of the geometry and of the fission product inventory. Some results are given and some improvements are planned

  10. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  11. Ultrasonic inspection for testing the PWR fuel rod endplug welds

    International Nuclear Information System (INIS)

    Pillet, C.; Destribats, M.T.; Papezyk, F.

    1976-01-01

    A method of ultrasonic testing with local immersion and transversal waves was developed. It is possible to detect defects as the lacks of fusion and penetration and porosity in the PWR fuel rod endplug welds [fr

  12. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  13. GO evaluation of a PWR spray system. Final report

    International Nuclear Information System (INIS)

    Long, W.T.

    1975-08-01

    GO is a reliability analysis methodology developed over the years from 1960 to the present by Kaman Sciences Corporation, Colorado Springs, Colorado. In this report the GO methodology is presented and its application demonstrated by performing a reliability analysis of a conceptual PWR Containment Spray System. Certain numerical results obtained are compared with those of a prior fault tree analysis of the same system as documented in the 11 January 1973 draft report, A Fault Tree Evaluation of a PWR Spray System

  14. Algorithms for the on-line travelling salesman

    NARCIS (Netherlands)

    Ausiello, G.; Feuerstein, E.; Leonardi, S.; Stougie, L.; Talamo, M.

    1999-01-01

    In this paper the problem of efficiently serving a sequence of requests presented in an on-line fashion located at points of a metric space is considered. We call this problem the On-Line Travelling Salesman Problem (OLTSP). It has a variety of relevant applications in logistics and robotics. We

  15. An algorithm for on-line price discrimination

    NARCIS (Netherlands)

    D.D.B. van Bragt; D.J.A. Somefun (Koye); E. Kutschinski; J.A. La Poutré (Han)

    2002-01-01

    textabstractThe combination of on-line dynamic pricing with price discrimination can be very beneficial for firms operating on the Internet. We therefore develop an on-line dynamic pricing algorithm that can adjust the price schedule for a good or service on behalf of a firm. This algorithm (a

  16. On-Line Voltage Stability Assessment based on PMU Measurements

    DEFF Research Database (Denmark)

    Garcia-Valle, Rodrigo; P. Da Silva, Luiz C.; Nielsen, Arne Hejde

    2009-01-01

    This paper presents a method for on-line monitoring of risk voltage collapse based on synchronised phasor measurement. As there is no room for intensive computation and analysis in real-time, the method is based on the combination of off-line computation and on-line monitoring, which are correlat...

  17. From Off-line to On-line Handwriting Recognition

    NARCIS (Netherlands)

    Lallican, P.; Viard-Gaudin, C.; Knerr, S.

    2004-01-01

    On-line handwriting includes more information on time order of the writing signal and on the dynamics of the writing process than off-line handwriting. Therefore, on-line recognition systems achieve higher recognition rates. This can be concluded from results reported in the literature, and has been

  18. On-line Learning of Prototypes and Principal Components

    NARCIS (Netherlands)

    Biehl, M.; Freking, A.; Hölzer, M.; Reents, G.; Schlösser, E.; Saad, David

    1998-01-01

    We review our recent investigation of on-line unsupervised learning from high-dimensional structured data. First, on-line competitive learning is studied as a method for the identification of prototype vectors from overlapping clusters of examples. Specifically, we analyse the dynamics of the

  19. On-line and bulk analysis for the resource industries

    International Nuclear Information System (INIS)

    Lim, C.S.; Sowerby, B.D.; Tickner, J.R.; Madsen, I.C.

    2001-01-01

    Nuclear techniques are the basis of many CSIRO on-line and bulk analysis systems that are now widely used in the mineral and energy industries. The continuous analysis and rapid response of these systems have led to improved control of mining, processing and blending operations. This paper reviews recent developments in neutron, gamma-ray and X-ray techniques for on-line and bulk analysis by CSIRO Minerals including neutron techniques for the on-conveyor belt determination of the composition of cement raw meal, the on-line analysis of composition in pyrometallurgical applications, the on-conveyor belt determination of ash in coal, and the rapid and accurate determination of gold in bulk laboratory samples. The paper also discusses a new gamma-ray technique for the on-line determination of ash in coal and the application of X-ray diffraction techniques for the on-line determination of mineralogy in the cement industry

  20. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  1. Field experience with advanced methods of on-line monitoring of water chemistry and corrosion degradation in nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Stellwag, B [Siemens AG Unternehmensbereich KWU, Erlangen (Germany); Aaltonen, P [Technical Research Centre of Finland, Espoo (Finland); Hickling, J [CML GmbH, Erlangen (Germany)

    1997-02-01

    Advanced methods for on-line, in-situ water chemistry and corrosion monitoring in nuclear power stations have been developed during the past decade. The terms ``on-line`` and ``in-situ`` characterize approaches involving continuous measurement of relevant parameters in high temperature water, preferably directly in the systems and components and not in removed samples at room temperature. This paper describes the field experience to-date with such methods in terms of three examples: (1) On-line chemistry monitoring of the primary coolant during shutdown of a Type WWER-440 PWR. (2) Redox and corrosion potential measurements in final feedwater preheaters and steam generators of two large KWU PWRs over several cycles of plant operation. (3) Real-time, in-situ corrosion surveillance inside the calundia vault of a CANDU reactor. The way in which water chemistry sensors and corrosion monitoring sensors complement each other is outlined: on-line, in-situ measurement of pH, conductivity and redox potential gives information about the possible corrosivity of the environment. Electrochemical noise techniques display signals of corrosion activity under the actual environmental conditions. A common experience gained from separate use of these different types of sensors has been that new and additional information about plants and their actual process conditions is obtained. Moreover, they reveal the intimate relationship between the operational situation and its consequences for the quality of the working fluid and the corrosion behaviour of the plant materials. On this basis, the efficiency of the existing chemistry sampling and control system can be checked and corrosion degradation can be minimized. Furthermore, activity buildup in the primary circuit can be studied. Further significant advantages can be expected from an integration of these various types of sensors into a common water chemistry and corrosion surveillance system. (Abstract Truncated)

  2. Understanding on-line community: the affordances of virtual space

    Directory of Open Access Journals (Sweden)

    Karen Ruhleder

    2002-01-01

    Full Text Available Increasing numbers of on-line venues for learning are emerging as virtual communities become more accessible and commonplace. This paper looks at one particular virtual community, an on-line degree programme at the University of Illinois, Urbana-Champaign, which offers an M.S. in Library and Information Science (called LEEP. It draws on a framework presented by Mynatt, et al. (1998, which provides a lens for talking about on-line community as a set of affordances. This framework is applied to illustrate the interactions, artefacts, and expectations that shape this community.

  3. An Approach to On-line Risk Assessment in NPP

    International Nuclear Information System (INIS)

    Simic, Z.; Mikulicic, V.; O'Brien, J.

    1996-01-01

    Probabilistic Risk Assessment (PRA) can provide safety status information for a plant during different configurations; additional effort is needed to do this in real time for on-line operation. This paper describes an approach to use PRA to achieve these goals. A Risk Assessment On-Line (RAOL) application was developed to monitor maintenance (on-line and planned) activities. RAOL is based on the results from a full-scope PRA, engineering/operational judgment and incorporates a user friendly program interface approach. Results from RAOL can be used by planners or operations to effectively manage the level of risk by controlling the actual plant configuration. (author)

  4. On-line spectroscopy with thermal atomic beams

    International Nuclear Information System (INIS)

    Thibault, C.; Guimbal, P.; Klapisch, R.; Saint Simon, M. de; Serre, J.M.; Touchard, F.; Duong, H.T.; Jacquinot, P.; Juncar, P.

    1981-01-01

    On-line high resolution laser spectroscopy experiments have been performed in which the light from a cw tunable dye laser interacts at right angles with a thermal atomic beam. sup(76-98)Rb, sup(118-145)Cs and sup(208-213)Fr have been studied using the ionic beam delivered by the ISOLDE on-line mass separator at CERN while sup(20-31)Na and sup(38-47)K have been studied by setting the apparaturs directly on-line with the PS 20 GeV proton beam. The principle of the method is briefly explained and some results concerning nuclear structure are given. (orig.)

  5. Comparison of On-Line Maintenance Support Tools

    International Nuclear Information System (INIS)

    Simic, Z.; Follen, S. M.; Mikulicic, V.

    1998-01-01

    Modeling approach to on-line risk monitoring is today in a rapid developing phase. For that reason number of different solutions are available. This paper will attempt to present existing approaches to address on-line risk modeling problem Starting with description of on-line risk monitoring issues in general, then following by presentation of existing software tools (EPRI's Safety Monitor, Equipment Out of Service Monitor, and ORAM-SENTINEL) the current state of the art in this area will be demonstrated. Finally, conclusions and ideas will be outlined. (author)

  6. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  7. Aging effects in PWR power plants components

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Diogo da S.; Guimaraes, Antonio C.F.; Moreira, Maria de Lourdes, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Fussell-Vesely Importance Measure. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of $27 million for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra I in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, if it had reached the time operating limit of forty years. The Extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the injection system of the containment spray) proposed in this work, it can be seen that 'the increase in the rate of component failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components'. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging process in Nuclear Plant Systems. (author)

  8. Maintenance technologies for SCC of PWR

    International Nuclear Information System (INIS)

    Okimura, Koji; Hori, Nobuyuki; Kanzaki, Hiroshi; Tokuhisa, Kiichi; Kamo, Kazuhiko; Kurokawa, Masaaki

    2007-01-01

    The recent technologies of test, relaxation of deterioration, repairing and change of materials are explained for safe and stable operation of pressurized water reactor (PWR). Stress corrosion cracking (SCC) is originated by three factors such as materials, stress and environment. The eddy current test (ECT) method for the stream generator pipe and the ultrasonic test method for welding part of pipe were developed as the test technologies. Primary water stress corrosion cracking (PWSCC) of Inconel 600 in the welding part is explained. The shot peening of instrument in the gas, the water jet peening of it in water, and laser irradiation on the surface are illustrated as some examples of improvement technology of stress. The cladding of Inconel 690 on Inconel 600 is carried out under the condition of environmental cut. Total or some parts of the upper part of reactor, stream generator and structure in the reactor are changed by the improvement technologies. Changing Inconel 600 joint in the exit pipe of reactor with Inconel 690 is illustrated. (S.Y.)

  9. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  10. Conceptual study on advanced PWR system

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

  11. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  12. Summary of PWR leak detection studies

    International Nuclear Information System (INIS)

    Cho, J.H.; Elia, F.A. Jr.

    1986-01-01

    Thermal-hydraulic analysis can be used to determine the location and magnitude of leaks inside and location of leaks outside a pressurized water reactor (PWR) containment as required by plant technical specifications. The major advantage of this detection method is that it minimizes radiation exposure of maintenance personnel because most of the leak detection process is performed from the control room outside containment. Plant-specific analyses are utilized to predict change in parameters such as local dew point temperature, relative humidity, dry bulb temperature, and flow rate to sump for various leak rates and enthalpies. These parameter responses are then programmed into the plant computer and instrumentation is provided for area monitoring. The actual inputs are continuously monitored and compared to the predicted plant responses to identify the leak location and quantify the leak. This study concludes that a system that monitors dew point (or relative humidity) and dry bulb temperature changes together with the flow rate to the sump will provide the capability to both locate and quantify a leak inside a containment, while a system that monitors dew point temperature (or relative humidity) changes will provide the capability to locate a leak outside a containment

  13. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1984-01-01

    Hydrogen generation during a PWR LOCA has been estimated for design basis accident and for two more severe hypothetical accidents. Hydrogen production during design basis accident is a rather slow mechanism, allowing in the worst case, 15 days to connect a hydrogen recombining unit to the containment atmosphere monitoring system. Hydrogen generated by steam oxidation during more severe hypothetical accidents was found limited by steam availability and fuel melting phenomena. Uncertainty is, however, still remaining on corium-zirconium-steam interaction. In the worst case, calculations lead to the production of 500 kg of hydrogen, thus leading to a volume concentration of 15% in containment atmosphere, assuming homogeneous hydrogen distribution within the reactor building. This concentration is within flammability limits but not within detonation limits. However, hydrogen detonation due to local hydrogen accumulation cannot be discarded. A major uncertainty subsisting on hydrogen hazard is hydrogen distribution during the first hours of the accident. This point determines the effects and consequences of local detonation or deflagration which could possibly be harmful to safeguard systems, or induce missile generation in the reactor building. As electrical supply failures are identified as an important contributor to severe accident risk, corrective actions have been taken in France to improve their reliability, including the installation of a gas turbine on each site to supplement the existing sources. These actions are thus contributing to hydrogen hazard reduction

  14. Analysis of reactivity accidents in PWR'S

    International Nuclear Information System (INIS)

    Camous, F.; Chesnel, A.

    1989-12-01

    This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

  15. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  16. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  17. Multicriteria analysis of public protection in PWR's

    International Nuclear Information System (INIS)

    Lombard, J.

    1986-09-01

    In order to manage a risk efficiently and to reach the ALARA level of protection, the best possible protection options must be employed. As the available resources are limited, it is not always possible to choose those options that minimize the risk, therefore a compromise must be made between risks and safety expenses. When the choice is difficult or complex, finding such a compromise can be facilitated by resorting to a decision aiding method which allows the assessment of the respective advantages of the various protection options considered. The multicriteria methods employ successive comparisons. Instead of searching for a final indicator expressing the performance of each option they compare all option pairs in order to determine if the gap between their respective advantages and disadvantages is sufficient to estimate that one option of the each pair is better than the other. Instead of judging each option globally these methods evaluate the advantages and disadvantages associated with the eventual choice of an option as compared with the others. These differential and comparative approach gives more flexibility and allows the introduction of qualitative criteria. The method presented here (Electre 3), one of the most recent ones, allows a multicriteria comparison of a set of options keeping into account the uncertainties associated with the options or the preferences. In order to illustrate this method a simple example (4 options, 4 criteria) dealing with a PWR liquid releases treatment system, is taken up

  18. Modeling of PWR fuel at extended burnup

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael M.; Silva, Antonio Teixeira, E-mail: rmdias@ipen.br, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  19. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  20. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  1. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

    1997-07-01

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  2. Computer aided information system for a PWR

    International Nuclear Information System (INIS)

    Vaidian, T.A.; Karmakar, G.; Rajagopal, R.; Shankar, V.; Patil, R.K.

    1994-01-01

    The computer aided information system (CAIS) is designed with a view to improve the performance of the operator. CAIS assists the plant operator in an advisory and support role, thereby reducing the workload level and potential human errors. The CAIS as explained here has been designed for a PWR type KLT- 40 used in Floating Nuclear Power Stations (FNPS). However the underlying philosophy evolved in designing the CAIS can be suitably adopted for other type of nuclear power plants too (BWR, PHWR). Operator information is divided into three broad categories: a) continuously available information b) automatically available information and c) on demand information. Two in number touch screens are provided on the main control panel. One is earmarked for continuously available information and the other is dedicated for automatically available information. Both the screens can be used at the operator's discretion for on-demand information. Automatically available information screen overrides the on-demand information screens. In addition to the above, CAIS has the features of event sequence recording, disturbance recording and information documentation. CAIS design ensures that the operator is not overburdened with excess and unnecessary information, but at the same time adequate and well formatted information is available. (author). 5 refs., 4 figs

  3. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  4. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael M.; Silva, Antonio Teixeira

    2015-01-01

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  5. ABB PWR fuel design for high burnup

    International Nuclear Information System (INIS)

    Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.

    1998-01-01

    Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)

  6. On-line methanol sensor system development for recombinant ...

    African Journals Online (AJOL)

    On-line methanol sensor system development for recombinant human serum ... of the methanol sensor system was done in a medium environment with yeast cells ... induction at a low temperature and a pH where protease does not function.

  7. On-line control systems in power plants

    International Nuclear Information System (INIS)

    Freymeyer, P.

    1981-01-01

    This report is a review of on-line control systems as a complex system connected with all problems like, development, planning, degree of automation, economics, service, quality and documentation. (orig.) [de

  8. Why do People Stop Playing On-Line Games?

    DEFF Research Database (Denmark)

    Sudzina, Frantisek; Razmerita, Liana

    2012-01-01

    The recent initial public offering of shares of Zynga, probably the most important on-line game provider, drew interest of potential investors but also of general public to their business model. What the most interested people learned so far is that if Zynga had not changed their accounting...... practice, they would be in red numbers for several months already. This is most likely caused by people stopping to play their games. This paper provides an estimate of what proportion of people, who played on-line games, already stopped playing them. Additionally, it analyzed the reasons why people...... stopped playing on-line games. It also compares Facebook and other on-line games....

  9. The Daresbury On-Line Isotope Separator (DOLIS)

    International Nuclear Information System (INIS)

    Grant, I.S.; Eastham, D.A.; Groves, J.; Tolfree, D.W.L.; Walker, P.M.; Green, V.R.; Rikovska, J.; Stone, N.J.; Hamilton, W.D.

    1987-01-01

    The isotope separator DOLIS, which is on-line to the Daresbury Laboratory's 20-MV tandem accelerator, is used to measure nuclear moments and decay schemes. Separated beams may be collected on a tape and transported to a counting station, implanted directly into a host lattice at on-line temperatures down to less than 10 mK, or allowed to interact with a collinear laser beam. The present status of DOLIS and its ancillary equipment is described

  10. The Daresbury on-line isotope separator (DOLIS)

    International Nuclear Information System (INIS)

    Grant, I.S.; Eastham, D.A.; Groves, J.; Tolfree, D.W.L.; Walker, P.M.; Green, V.R.; Rikovska, J.; Stone, N.J.; Hamilton, W.D.

    1987-01-01

    The isotope separator DOLIS, which is on-line to the Daresbury Laboratory's 20-MV tandem accelerator, is used to measure nuclear moments and decay schemes. Separated beams may be collected on a tape and transported to a counting station, implanted directly into a host lattice at on-line temperatures down to less than 10 mK, or allowed to interact with a collinear laser beam. The present status of DOLIS and its ancillary equipment is described. (orig.)

  11. Considerations in applying on-line IC techniques to BWR's

    International Nuclear Information System (INIS)

    Kaleda, R.J.

    1992-01-01

    Ion-Chromatography (IC) has moved from its traditional role as a laboratory analytical tool to a real time, dynamic, on-line measurement device to follow ppb and sub-ppb concentrations of deleterious impurities in nuclear power plants. Electric Power Research Institute (EPRI), individual utilities, and industry all have played significant roles in effecting the transition. This paper highlights considerations and the evolution in current on-line Ion Chromatography systems. The first applications of on-line techniques were demonstrated by General Electric (GE) under EPRI sponsorship at Rancho Seco (1980), Calvert Cliffs, and McGuire nuclear units. The primary use was for diagnostic purposes. Today the on-line IC applications have been expanded to include process control and routine plant monitoring. Current on-line IC's are innovative in design, promote operational simplicity, are modular for simplified maintenance and repair, and use field-proven components which enhance reliability. Conductivity detection with electronic or chemical suppression and spectrometric detection techniques are intermixed in applications. Remote multi-point sample systems have addressed memory effects. Early applications measured ionic species in the part per billion range. Today reliable part per trillion measurements are common for on-line systems. Current systems are meeting the challenge of EPRI guideline requirements. Today's on-line IC's, with programmed sampling systems, monitor fluid streams throughout a power plant, supplying data that can be trended, stored and retrieved easily. The on-line IC has come of age. Many technical challenges were overcome to achieve today's IC

  12. On-line mixture-based alternative to logistic regression

    Czech Academy of Sciences Publication Activity Database

    Nagy, Ivan; Suzdaleva, Evgenia

    2016-01-01

    Roč. 26, č. 5 (2016), s. 417-437 ISSN 1210-0552 R&D Projects: GA ČR GA15-03564S Institutional support: RVO:67985556 Keywords : on-line modeling * on-line logistic regression * recursive mixture estimation * data dependent pointer Subject RIV: BB - Applied Statistics, Operational Research Impact factor: 0.394, year: 2016 http://library.utia.cas.cz/separaty/2016/ZS/suzdaleva-0464463.pdf

  13. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  14. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  15. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  16. Overview of the Vercors programme devoted to safety studies on irradiated PWR fuel

    International Nuclear Information System (INIS)

    Tourasse, M.; Andre, B.; Ducros, G.; Maro, D.

    1996-01-01

    The first objective of the Heva-Vercors Program is to improve the data base of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EdF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation : gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on-line and post-test characterization. The knowledge of the behaviour of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. (author)

  17. Overview of the Vercors Programme Devoted to Safety Studies on Irradiated PWR Fuel

    International Nuclear Information System (INIS)

    Tourasse, M.; Andre, B.; Ducros, G.; Maro, D.

    1996-01-01

    The first objective of the Heva-Vercors Program is to improve the data of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EDF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation: gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on line and post test characterization. The knowledge of the behavior of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. The first aim of these future tests is to study the behaviour of non volatile and transuranic elements. An even more sophisticated instrumentation is implemented to reach the goal. The use of MOX fuel, the interaction between fission product aerosols and structural materials (Ag-In-Cd) and the fuel granulometry effect will be the next steps of the experimental program

  18. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  19. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  20. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  1. Basic information about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) Plant layout of a PWR; 2.0) principle design of a PWR and the reactor coolant system; 3.0) reactor auxiliary and ancillary systems; 3.1) volume control system; 3.2) boric acid control and chemical feeding system; 3.3) coolant purification and degassing system; 3.4) coolant storage and treatment system; 3.5) nuclear component cooling system; 3.6) liquid waste processing system; 3.7) gaseous waste processing system; 4.0) residual heat removal system; 5.0) emergency feedwater system; 6.0) containment design; 7.0) fuel handling, storage and transport system in a PWR. (orig.) [de

  2. Improved emergency elevated air release for simplified PWR

    International Nuclear Information System (INIS)

    Naitoh, T.; Bruce, R.A.; Hirota, K.; Tajiri, Y.

    1992-01-01

    In developing the application of the simplified PWR in Japan, one of the most important areas is to limit post-accident site boundary whole body dose. In addressing this, the concept of Emergency Passive Air Filtration System (EPAFS) and it's feasibility is developed. The efficiency of charcoal filtering and the atmospheric diffusion effect of an elevated air release are important for dose reduction. The performance of these functions was evaluated by confirmatory testing. The test results confirmed a 99 percent efficiency of charcoal filter and an atmospheric diffusion effect higher than that of a conventional plant. The Emergency Passive Air Filtration System (EPAFS) and the atmospheric diffusion effect of elevated air release contribute to making the calculated post-accident site boundary whole body dose of simplified PWR as low as that of the conventional Japanese PWR plant. (author)

  3. Swing-Down of 21-PWR Waste Package

    International Nuclear Information System (INIS)

    A.K. Scheider

    2001-01-01

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

  4. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  5. Fabrication of PWR fuel assembly and CANDU fuel bundle

    International Nuclear Information System (INIS)

    Lee, G.S.; Suh, K.S.; Chang, H.I.; Chung, S.H.

    1980-01-01

    For the project of localization of nuclear fuel fabrication, the R and D to establish the fabrication technology of CANDU fuel bundle as well as PWR fuel assembly was carried out. The suitable boss height and the prober Beryllium coating thickness to get good brazing condition of appendage were studied in the fabrication process of CANDU fuel rod. Basic Studies on CANLUB coating method also were performed. Problems in each fabrication process step and process flow between steps were reviewed and modified. The welding conditions for top and bottom nozzles, guide tube, seal and thimble screw pin were established in the fabrication processes of PWR fuel assembly. Additionally, some researches for a part of PWR grid brazing problems are also carried out

  6. The advanced main control console for next japanese PWR plants

    International Nuclear Information System (INIS)

    Tsuchiya, A.; Ito, K.; Yokoyama, M.

    2001-01-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  7. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  8. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

  9. Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

    International Nuclear Information System (INIS)

    Fairclough, M.P.; Tymons, B.J.

    1985-06-01

    The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

  10. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  11. On-line maintenance PSA support at NPP Krsko

    International Nuclear Information System (INIS)

    Prosen, R.; Vrbanic, I.; Kastelan, M.

    2000-01-01

    In 1997 Krsko NPP initiated the on-line maintenance (OLM) practice. On-line maintenance constitutes of corrective activities, preventive activities, surveillance activities, tests and inspections, as well as calibrations and modifications, taking place during the normal power operations. The on-line maintenance is a multidisciplinary process consisting of activity specification, planning, and preparation and performing of the OLM activity of interest. The primary role of the PSA group is to assess from the r isk perspective , using the plant-specific NEK PSA model, system unavailability and the impact to the plant operational risk. The intent is to support planning of the on-line maintenance activities from the risk perspective. The risk evaluation of the OLM activities is based on the probability of core damage evaluation for the defined discrete plant configuration states, determined by the OLM activities. Within this application, the optimized, plant-specific PSA model is used on Risk Spectrum platform. To perform the risk assessment of the on-line maintenance activities, first the systems to be affected are defined based on the planned OLM activities. The next important step is the assessment of the planned work schedule. To define the final schedule, the co-ordination and optimizing the planned OLM activities needs activation of all participating departments, supported also from PSA group. The P3 (i.e. Primavera) planning tool system windows are defined for different systems and groups of systems, and the activities are sorted in particular weeks according to these windows. (author)

  12. OLDASS: On-line data acquisition system at SF cyclotron

    International Nuclear Information System (INIS)

    Omata, Kazuo; Yasue, Masaharu; Hamagaki, Hideki

    1982-01-01

    The on-line data acquisition system in the Institute for Nuclear Study, the University of Tokyo, is composed of 2 systems of Fujitsu mini-computer PFU-400 for data processing at the high energy synchrotron and one system of that computer for low energy cyclotron as terminals, the host computer being M 180 II AD of the same company. This system has been developed to have the features of being the on-line system capable of following the improvement of host computer performance, being capable of developing the on-line programmes of other experimenting groups in parallel with batch jobs or the operation of the on-line system, and capable of developing programmes using FORTRAN. The result of about 220 KB/s was obtained for the data transfer rate between the programmes of the host computer and terminals, and this fulfilled the aimed performance. The terminal system on the low energy side is provided with an ADC interface and a display interface specified particularly in addition to the miniature computer PFU400 and standard I/O devices of the manufacture. The accumulating type graphic display of the I/O devices can be switched to be connected to the host computer, and immediately displays the results transferred to the host computer and analyzed. Hard copy is also available. The above hardware and software are explained. The on-line system insures 80 K bytes of the total memory of 224 K bytes for data area. (Wakatsuki, Y.)

  13. Sport Management Taught On-Line: A Discussion

    Directory of Open Access Journals (Sweden)

    William F. Stier Jr

    2009-01-01

    Full Text Available An introduction to the world of on-line courses (distance education/learning is presented. In addition, the world of on-line learning, as it pertains to sport management, is examined within the framework of (a pedagogy, (b finances,(c assessment, and (d choosing to transition from the traditional classroom to on-line learning. Pertinent points relative to each of the four categories are presented from the literature. In an effort to stimulate thought and discussion to the subject of on-line learning for sport management programs/courses the authors provide their reactions to the literature points by presenting their comments/reactions from a sport management perspective. Sport management professors and administrators are encouraged to critically examine the feasibility of such on-line courses (distance education/learning within their own curricula while maintaining an appropriate framework revolving around sound theoretical instructional strategies, methods as well as appropriate use of instructional tools, including but not limited to, computersand the WWW.

  14. The traveller: a new look for PWR fresh fuel packages

    International Nuclear Information System (INIS)

    Bayley, B.; Stilwell, W.E.; Kent, N.A.

    2004-01-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. This paper follows the development effort from the establishment of goals and objectives, to intermediate testing and analysis, to final testing and licensing. The discussion starts with concept origination and covers the myriad iterations that followed until arriving at a design that would meet the demanding licensing requirements, last for 30 years, and would be easy to load and unload fuel, easy to handle, inexpensive to manufacture and transport, and simple and inexpensive to maintain

  15. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  16. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  17. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  19. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

  20. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  1. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  2. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    Hoffman, B.; Tsuzuki, S.

    2002-01-01

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  3. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  4. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  5. Leak before break application in French PWR plants under operation

    Energy Technology Data Exchange (ETDEWEB)

    Faidy, C. [EDF SEPTEN, Villeurbanne (France)

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  6. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  7. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  8. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  9. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  10. Initiating Events Modeling for On-Line Risk Monitoring Application

    International Nuclear Information System (INIS)

    Simic, Z.; Mikulicic, V.

    1998-01-01

    In order to make on-line risk monitoring application of Probabilistic Risk Assessment more complete and realistic, a special attention need to be dedicated to initiating events modeling. Two different issues are of special importance: one is how to model initiating events frequency according to current plant configuration (equipment alignment and out of service status) and operating condition (weather and various activities), and the second is how to preserve dependencies between initiating events model and rest of PRA model. First, the paper will discuss how initiating events can be treated in on-line risk monitoring application. Second, practical example of initiating events modeling in EPRI's Equipment Out of Service on-line monitoring tool will be presented. Gains from application and possible improvements will be discussed in conclusion. (author)

  11. Strengthening weak ties through on-line gaming

    DEFF Research Database (Denmark)

    Sudzina, Frantisek; Razmerita, Liana Virginia; Kirchner, Kathrin

    On-line gaming became widespread in the last couple of years. The aim of the research presented in the paper is to figure out to what extent does game playing helps to strengthen weak ties and what additional factors influence this process. The approach is rather exploratory – some factors...... are grounded in theory, some are new. These factors are age, gender, place of origin, number of their Facebook connections (friends in Facebook terminology), the amount of time they are on Facebook, the amount of time they keep the Facebook site open, the amount of time they play on-line games, and the reasons...... for starting to play on-line games. Regarding the latter, we chose to focus only on escapist reasons....

  12. Thermographic Sensing For On-Line Industrial Control

    Science.gov (United States)

    Holmsten, Dag

    1986-10-01

    It is today's emergence of thermoelectrically cooled, highly accurate infrared linescanners and imaging systems that has definitely made on-line Infraread Thermography (IRT) possible. Specifically designed for continuous use, these scanners are equipped with dedicated software capable of monitoring and controlling highly complex thermodynamic situations. This paper will outline some possible implications of using IRT on-line by describing some uses of this technology in the steel-making (hot rolling) and automotive industries (machine-vision). A warning is also expressed that IRT technology not originally designed for automated applications e.g. high resolution, imaging systems, should not be directly applied to an on-line measurement situation without having its measurement resolution, accuracy and especially its repeatability, reliably proven. Some suitable testing procedures are briefly outlined at the end of the paper.

  13. An on-line monitoring system for navigation equipment

    Science.gov (United States)

    Wang, Bo; Yang, Ping; Liu, Jing; Yang, Zhengbo; Liang, Fei

    2017-10-01

    Civil air navigation equipment is the most important infrastructure of Civil Aviation, which is closely related to flight safety. In addition to regular flight inspection, navigation equipment's patrol measuring, maintenance measuring, running measuring under special weather conditions are the important means of ensuring aviation flight safety. According to the safety maintenance requirements of Civil Aviation Air Traffic Control navigation equipment, this paper developed one on-line monitoring system with independent intellectual property rights for navigation equipment, the system breakthroughs the key technologies of measuring navigation equipment on-line including Instrument Landing System (ILS) and VHF Omni-directional Range (VOR), which also meets the requirements of navigation equipment ground measurement set by the ICAO DOC 8071, it provides technical means of the ground on-line measurement for navigation equipment, improves the safety of navigation equipment operation, and reduces the impact of measuring navigation equipment on airport operation.

  14. On-line calculation of 3-D power distribution

    International Nuclear Information System (INIS)

    Park, Y. H.; In, W. K.; Park, J. R.; Lee, C. C.; Auh, G. S.

    1996-01-01

    The 3-D power distribution synthesis scheme was implemented in Totally Integrated Core Operation Monitoring System (TICOMS), which is under development as the next generation core monitoring system. The on-line 3-D core power distribution obtained from the measured fixed incore detector readings is used to construct the hot pin power as well as the core average axial power distribution. The core average axial power distribution and the hot pin power of TICOMS were compared with those of the current digital on-line core monitoring system, COLSS, which construct the core average axial power distribution and the pseudo hot pin power. The comparison shows that TICOMS results in the slightly more accurate core average axial power distribution and the less conservative hot pin power. Therefore, these results increased the core operating margins. In addition, the on-line 3-D power distribution is expected to be very useful for the core operation in the future

  15. Approaches for on-line coupling of extraction and chromatography

    Energy Technology Data Exchange (ETDEWEB)

    Hyoetylaeinen, Tuulia; Riekkola, Marja-Liisa [Laboratory of Analytical Chemistry, Department of Chemistry, University of Helsinki, P.O. Box 55, 00014, Helsinki (Finland)

    2004-04-01

    This review provides an overview of the approaches available in order to perform on-line coupling of various extraction techniques with liquid and gas chromatography, for the analysis of semivolatile and nonvolatile analytes in liquid and solid samples. The main focus is on the instrumental set-up of these techniques. Selected real applications are described by way of illustration. The extraction methods suitable for on-line coupling covered in this review are: liquid-liquid extraction, solid-phase extraction, membrane-based techniques, pressurised liquid extraction, supercritical fluid extraction, and microwave- and sonication-assisted extractions. The following systems are not covered in this review: on-line coupled solid-phase extraction-liquid chromatography, purge-and-trap-GC, and membrane extraction with a sorbent interface-GC. (orig.)

  16. UNISOR on-line nuclear orientation facility (UNISOR/NOF)

    International Nuclear Information System (INIS)

    Girit, I.C.; Alton, G.D.; Bingham, C.R.; Carter, H.K.; Simpson, M.L.; Cole, J.D.; Croft, W.L.; Hamilton, J.H.; Jones, E.F.; Gore, P.M.; Kormicki, J.; Xie, H.; Kern, B.D.; Krane, K.S.; Xu, Y.S.; Mantica, P.F. Jr.; Zimmermann, B.E.; Nettles, W.G.; Zganjar, E.F.; Kortelahti, M.O.; Newbolt, W.B.

    1988-01-01

    The UNISOR on-line nuclear orientation facility (UNISOR/NOF) consists of a 3 He- 4 He dilution refrigerator on line to the isotope separator. Nuclei are implanted directly into a target foil which is soldered to the bottom accessed cold finger of the refrigerator. A 1.5 T superconducting magnet polarizes the ferromagnetic target foils and determines the axis of symmetry. Up to eight gamma detectors can be positioned around the refrigerator, each 9 cm from the target. A unique feature of this system is that the k=4 term in the directional distribution function can be directly and unambigously deduced so that a single solution for the mixing ratio can be found. The first on-line experiment at this facility reported here was a study of the decay of the 191 Hg and 193 Hg isotopes. (orig.)

  17. Making the most of on-line recruiting.

    Science.gov (United States)

    Cappelli, P

    2001-03-01

    Ninety percent of large U.S. companies are already recruiting via the Internet. By simply logging on to the Web, company recruiters can locate vast numbers of qualified candidates for jobs at every level, screen them in minutes, and contact the most promising ones immediately. The payoffs can be enormous: it costs substantially less to hire someone on-line, and the time saved is equally great. In this article, Peter Cappelli examines some of the emerging service providers and technologies--matchmakers, job boards, hiring management systems software, and applicant-screening mechanisms that test skills and record interests. He also looks at some of the strategies companies are adopting as they enter on-line labor markets. Recruiting needs to be refashioned to resemble marketing, he stresses. Accordingly, smart companies are designing Web pages, and even product ads, with potential recruits in mind. They're giving line managers authority to hire so that candidates in cyberspace aren't lost. They're building internal on-line job networks to retain talent. Integrating recruiting efforts with overall marketing campaigns, especially through coordination and identification with the company's brand, is the most important thing companies can do to ensure success in on-line hiring. Along the way, Cappelli sounds two cautionary notes. First, a human touch, not electronic contact, is vital in the last steps of a successful hiring process. Second, companies must make sure that on-line testing and hiring criteria do not discriminate against women, disabled people, workers over 40, or members of minority groups. When competition for talent is fierce, companies that master the art and science of on-line recruiting will be the ones that attract and keep the best people.

  18. The User-friendly On-Line Diffusion Chamber

    CERN Document Server

    Aviles Acosta, Jaime

    2015-01-01

    The On-Line Diffusion Chamber is a stand-alone apparatus built to carry out short-live radiotracer diffusion studies. The availability of the on-demand production of isotopes in the ISOLDE facility, and the design of the apparatus to streamline the implantation process, annealing treatment, ion gun ablation with a tape transport system, and radiation intensity measurement with a Ge gamma detector all in the same apparatus, gives the On-Line Diffusion Chamber a unique ability to studies with short-lived radioisotopes or isomer states that are not possible in any other facility in the world.

  19. The rationalization of desulfurization by on-line analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Y; Kohmura, S; Taketomi, H; Matsumura, S; Sasaki, Y

    1986-01-01

    Nippon Kokan uses the Takahax and Sulfiban processes for the desulfurization of coke oven gas. The authors outline the Sulfiban Process and describe a recently developed system for the on-line determination of H/sub 2/S in coke oven gas, and of CO/sub 2/ and monoethanolamine (MEA) in the wash oil. This new on-line analysis system has proved effective in rationalizing the Sulfiban Process via lower MEA production costs and decreased power consumption. The introduction of a computerized control system is now being studied. 7 figs., 4 tabs.

  20. 5th Computer Science On-line Conference

    CERN Document Server

    Senkerik, Roman; Oplatkova, Zuzana; Silhavy, Petr; Prokopova, Zdenka

    2016-01-01

    This volume is based on the research papers presented in the 5th Computer Science On-line Conference. The volume Artificial Intelligence Perspectives in Intelligent Systems presents modern trends and methods to real-world problems, and in particular, exploratory research that describes novel approaches in the field of artificial intelligence. New algorithms in a variety of fields are also presented. The Computer Science On-line Conference (CSOC 2016) is intended to provide an international forum for discussions on the latest research results in all areas related to Computer Science. The addressed topics are the theoretical aspects and applications of Computer Science, Artificial Intelligences, Cybernetics, Automation Control Theory and Software Engineering.

  1. Comparison of Current On-line Payment Technologies

    OpenAIRE

    Mandadi, Ravi

    2006-01-01

    The purpose of this thesis work was to make a survey of current on-line payment technologies and find out which are they and how do they work? Compare and analyze them from a security point of view, as well as a usability point of view. What is good? What is bad? What is lacking? To achieve this purpose, an overview of the current on-line payment technologies was acquired through academic books and papers, Internet sites, magazines. Basic cryptographic and security related techniques were stu...

  2. An on-line adaptive core monitoring system

    International Nuclear Information System (INIS)

    Verspeek, J.A.; Bruggink, J.C.; Karuza, J.

    1997-01-01

    An on-line core monitoring system has been in operation for three years in the Dodewaard Nuclear Power Plant. The core monitor uses the on-line measured reactor data as an input for a power distribution calculation. The measurements are frequently performed. The system is used for monitoring as well as for predicting purposes. The limiting thermal hydraulic parameters are monitored as well as the pellet-clad interaction limits. The data are added to a history file used for cycle burn-up calculations and trending of parameters. The reactor states are presented through a convenient graphical user interface. (authors)

  3. On-Line Generation of 3D-Waves

    DEFF Research Database (Denmark)

    Frigaard, Peter

    1992-01-01

    The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg.......The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg....

  4. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  5. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  6. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  7. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  8. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    Jaime, Guilherme D.G.; Oliveira, Mauro V.

    2011-01-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  9. The application of transition metal ion chromatography to the determination of elemental and radiochemical species in PWR primary coolant

    International Nuclear Information System (INIS)

    Bridle, D.A.; Brown, G.R.; Johnson, P.A.V.

    1992-01-01

    The accurate determination of both elemental and radiochemical transition metal corrosion products, particularly cobalt and nickel, in PWR coolants is necessary if the transport mechanisms and their role in the development of out-of-core radiation fields are to be fully understood. AEA Technology, Winfrith, has collaborated for several years with a number of PWR utilities in Europe, developing advanced sampling and analytical techniques for the determination of both soluble and insoluble corrosion products in primary coolant. The design and installation of continuously flowing isokinetic capillary modifications to the existing sampling systems has been shown to be an effective method of providing a low, but representative, sample flow from high pressure systems for on-line determination of corrosion product species. Transition metal ion chromatography coupled with gamma-spectrometry has been used to determine both insoluble and soluble elemental and radiochemical species in reactor coolant, with particular attention being given to the determination of soluble elemental cobalt at levels as low as 1 ng per kg. Soluble species were determined directly following their concentration from up to 1 litre of coolant. Insoluble species collected on 0.45 micron filter membranes, following filtration of up to 1500 litres of coolant, were solubilised by fusion with potassium hydrogen sulphate before the application of ion chromatography. In each case the eluant from the chromatographic column was collected and the radionuclides determined by gamma-spectrometry

  10. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

  11. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  12. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  13. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  14. Method of characteristics - Based sensitivity calculations for international PWR benchmark

    International Nuclear Information System (INIS)

    Suslov, I. R.; Tormyshev, I. V.; Komlev, O. G.

    2013-01-01

    Method to calculate sensitivity of fractional-linear neutron flux functionals to transport equation coefficients is proposed. Implementation of the method on the basis of MOC code MCCG3D is developed. Sensitivity calculations for fission intensity for international PWR benchmark are performed. (authors)

  15. Studies of a small PWR for onsite industrial power

    International Nuclear Information System (INIS)

    Klepper, O.H.; Smith, W.R.

    1977-01-01

    Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application

  16. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits [fr

  17. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  18. Dissolution process for advanced-PWR-type fuels

    International Nuclear Information System (INIS)

    Black, D.E.; Decker, L.A.; Pearson, L.G.

    1979-01-01

    The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

  19. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  20. Is it possible to improve regulation system of PWR

    International Nuclear Information System (INIS)

    Bonnemay, A.; Martinez, J.M.

    1983-03-01

    This paper deals with two problems: first of all, it presents the critical analysis of usually implemented general regulation systems, on PWR plants, and derives from it same possibilities to improve the transient behavior of reactor, the second part is a proposition from an automatic control system for spatial distribution of flux

  1. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

  2. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  3. Performance of PWR Nuclear power plants, up to 1985

    International Nuclear Information System (INIS)

    Muniz, A.A.

    1987-01-01

    The performance of PWR nuclear power plants is studied, based on operational data up to 1985. The availability analysis was made with 793 unit-year and the reliability analysis was made with 5851 unit x month. The results were discussed and the availability of those nuclear power plants were estimated. (E.G.) [pt

  4. PWR auxiliary systems, safety and emergency systems, accident analysis, operation

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1976-01-01

    The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK) [de

  5. The new French code for PWR in service inspection

    Energy Technology Data Exchange (ETDEWEB)

    Noel, R; Hutin, J P [Electricite de France (EDF), 75 - Paris (France)

    1988-12-31

    This document presents the new french code for pressured water reactor in service inspection. The historic regulatory basis is presented, together with the new regulatory act (dating back to the 26 february 1974) and the major guidelines of the french practice for in service inspection of PWR components. (TEC).

  6. Influence of boron reduction strategies on PWR accident management flexibility

    International Nuclear Information System (INIS)

    Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

    2007-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

  7. Directives and general design requirements for a small PWR

    International Nuclear Information System (INIS)

    Arrieta, L.A.

    1992-08-01

    A proposal of directives and general requirements for the development of a small PWR conceptual design is presented. These directives address the main safety, performance and economic design aspects. The purpose is to use this work as a base for a wide discussion, involving all project participants, culminating with the definition of the final directives and general requirements. (author)

  8. Post irradiation examination on test fuel pins for PWR

    International Nuclear Information System (INIS)

    Fogaca Filho, N.; Ambrozio Filho, F.

    1981-01-01

    Certain aspects of irradiation technology on test fuel pins for PWR, are studied. The results of post irradiation tests, performed on test fuel pins in hot cells, are presented. The results of the tests permit an evaluation of the effects of irradiation on the fuel and cladding of the pin. (Author) [pt

  9. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Pitts, M.L.

    2000-01-01

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

  10. Integrated sensor array for on-line monitoring micro bioreactors

    NARCIS (Netherlands)

    Krommenhoek, E.E.

    2007-01-01

    The “Fed��?batch on a chip��?��?project, which was carried out in close cooperation with the Technical University of Delft, aims to miniaturize and parallelize micro bioreactors suitable for on-line screening of micro-organisms. This thesis describes an electrochemical sensor array which has been

  11. On-Line Pesticide Training with Narrated Powerpoint Presentations

    Science.gov (United States)

    Johnson, Steven B.

    2015-01-01

    UMaine Cooperative Extension is the primary educational delivery organization for pesticide recertification credits in Maine. Shrinking budgets and staff numbers are making traditional face-to-face delivery increasingly difficult to maintain. To address this issue, on-line pesticide applicator recertification training credits were developed. The…

  12. Personality Interactions and Scaffolding in On-Line Discussions

    Science.gov (United States)

    Nussbaum, E. Michael; Hartley, Kendall; Sinatra, Gale M.; Reynolds, Ralph E.; Bendixen, Lisa D.

    2004-01-01

    The potential of on-line discussions to prompt greater reflection of course material is often stymied by a tendency of students to agree with one another rather than to formulate counter-arguments. This article describes an experiment using note starters and elaborated cases to encourage counter-argumentation and examines interactions with…

  13. A new electrostatic on-line collection-system

    International Nuclear Information System (INIS)

    Dufour, J.P.; Del Moral, R.; Fleury, A.; Hubert, F.; Llabador, Y.; Mauhourat, M.B.; Bimbot, R.; Gardes, D.; Rivet, M.F.

    1981-01-01

    The working conditions of a new on-line electrostatic collection system are presented. The main characteristics are high efficiency (reaching 20%) and short delay time (down to the millisecond). The salient features of specific devices for measurements of absolute cross sections, recoil range distributions and angular distributions are given. (orig.)

  14. Specialization processes in on-line unsupervised learning

    NARCIS (Netherlands)

    Biehl, M.; Freking, A.; Reents, G.; Schlösser, E.

    1998-01-01

    From the recent analysis of supervised learning by on-line gradient descent in multilayered neural networks it is known that the necessary process of student specialization can be delayed significantly. We demonstrate that this phenomenon also occurs in various models of unsupervised learning. A

  15. A simulation of the SDC on-line processing farm

    International Nuclear Information System (INIS)

    Wang, C.; Chen, Y.; Dorenbosch, J.; Lee, J.; Sayle, R.

    1993-10-01

    In the Solenoidal Detector Collaboration (SDC) data acquisition system (DAQ), an enormous amount of data flows into a processor farm for extraction of interesting physics events. To design an efficient on-line filter, the operations in the farm must be carefully modeled. The authors present a simulation model developed at the Superconducting Super Collider Laboratory which efficiently allocates physics events to the farm

  16. On-Line Synthesis and Analysis by Mass Spectrometry

    Science.gov (United States)

    Bain, Ryan M.; Pulliam, Christopher J.; Raab, Shannon A.; Cooks, R. Graham

    2015-01-01

    In this laboratory experiment, students learn how to use ESI to accelerate chemical synthesis and to couple it with on-line mass spectrometry for structural analysis. The Hantzsch synthesis of symmetric 1,4-dihydropyridines is a classic example of a one-pot reaction in which multiple intermediates can serve to indicate the progress of the reaction…

  17. An optimal algorithm for preemptive on-line scheduling

    NARCIS (Netherlands)

    Chen, B.; Vliet, van A.; Woeginger, G.J.

    1995-01-01

    We investigate the problem of on-line scheduling jobs on m identical parallel machines where preemption is allowed. The goal is to minimize the makespan. We derive an approximation algorithm with worst-case guarantee mm/(mm - (m - 1)m) for every m 2, which increasingly tends to e/(e - 1) ˜ 1.58 as m

  18. Comparison between constant methanol feed and on-line ...

    African Journals Online (AJOL)

    Two methanol feeding methods, namely constant methanol feed and on-line monitoring feed control by methanol sensor were investigated to improve the production of recombinant human growth hormone (rhGH) in high cell density cultivation of Pichia pastoris KM71 in 2 L bioreactor. The yeast utilized glycerol as a carbon ...

  19. Children's On-Line Processing of Scrambling in Japanese

    Science.gov (United States)

    Suzuki, Takaaki

    2013-01-01

    This study investigates the on-line processing of scrambled sentences in Japanese by preschool children and adults using a combination of self-paced listening and speeded picture selection tasks. The effects of a filler-gap dependency, reversibility, and case markers were examined. The results show that both children and adults had difficulty in…

  20. AAEC INIS - a large, new, on-line information source

    International Nuclear Information System (INIS)

    Rugg, T.J.; Wong, S.C.

    1984-01-01

    The Australian Atomic Energy Commission's INIS database is available for on-line searching by non-AAEC personnel from all parts of Australia. An introduction to the International Nuclear Information System is followed by information on searching AAEC INIS, AAEC INIS retrieval software and accessing AAEC INIS

  1. ORION-the Omega Remote Interactive On-line System

    CERN Document Server

    Russell, R D; Levratt, B; Lipps, H; Sparrman, P

    1974-01-01

    ORION is a system which permits the manipulation of files, records and characters, remote job submittal and retrieval of output files including the direct loading of remote on-line computers. The system uses the computer hardware of the OMEGA project at CERN and is designed to assist researchers in development and debugging of their programs. (10 refs).

  2. ORION - the OMEGA Remote Interactive On-line System

    CERN Document Server

    Russell, R D; Krieger, M

    1973-01-01

    ORION is a system which permits the manipulation of files, records and characters, remote job submittal and retrieval of output files including the direct loading of remote on-line computers. The system uses the computer hardware of the OMEGA project at CERN, and is designed to assist researchers in development and debugging of their programs.

  3. Booth Library On-Line Circulation System (BLOC

    Directory of Open Access Journals (Sweden)

    Paladugu V. Rao

    1971-06-01

    Full Text Available An on-line circulation system developed at a relatively small university library demonstrates that academic libraries with limited funds can develop automated systems utilizing parent institution's computer facilities in a time-sharing mode. In operation since September 1968, using an IBM 360/50 computer and associated peripheral equipment, it provides control over all stack books.

  4. On-line learning from clustered input examples

    NARCIS (Netherlands)

    Riegler, Peter; Biehl, Michael; Solla, Sara A.; Marangi, Carmela; Marinaro, Maria; Tagliaferri, Roberto

    1996-01-01

    We analyse on-line learning of a linearly separable rule with a simple perceptron. Example inputs are taken from two overlapping clusters of data and the rule is defined through a teacher vector which is in general not aligned with the connection line of the cluster centers. We find that the Hebb

  5. Investigating on-line pornography at the University of Johannesburg

    Directory of Open Access Journals (Sweden)

    P. Laughton

    2008-01-01

    Full Text Available The on-line user of today has access to a vast collection of information resources. In addition, the developments in Internet and Web technologies have made it even easier for surfers to anonymously get access to on-line pornography. The purpose of this research was to investigate the extent to which access to on-line pornography at the University of Johannesburg can be managed. For the empirical part of this research 1037 questionnaires were proportionally distributed to and completed by students on all five campuses of the university. The questionnaire consisted of four sections: biographical information; university computer facility usage; university acceptable use policy; and personal experience with university computer facilities. The gender distribution for the sample was almost even, with a total of 49,4% male participants and 50,6% female, with the largest grouping of respondents (61,6% aged between 19 years and 21 years. Of the respondents, 36,7% indicated that exposure to unsolicited pornography did not bother them. When asked to what extent students should have access to pornography, 60,5% stated 'None' while 32,6% believed that 'Restricted' access should be granted for research purposes and 6,9% believed that students should be granted 'Total' access to pornography. Results from the research will be used to manage access to on-line resources at the University of Johannesburg better.

  6. On-line sample treatment - Capillary gas chromatography

    NARCIS (Netherlands)

    Goosens, EC; de Jong, D; de Jong, GJ; Brinkman, UAT

    Sample pretreatment is often the bottleneck of a trace level analytical procedure. In order to increase performance, increasing attention is therefore being devoted to combining sample pretreatment on-line with the separation technique that has to be used. In the present review, a variety of

  7. A new electrostatic on-line collection-system

    International Nuclear Information System (INIS)

    Dufour, J.P.; Del Moral, R.; Fleury, A.

    1981-06-01

    The working conditions of a new on-line electrostatic collection system are presented. The main charactersitics are high efficiency (reaching 20%) and short delay time (down to the millisecond). The salient features of specific devices for measurements of absolute cross sections, recoil range distributions and angular distributions are given

  8. Summary remarks and prospects for on-line nuclear orientation

    International Nuclear Information System (INIS)

    Krane, K.S.; Hamilton, J.H.

    1984-01-01

    This paper reviews the use of on-line nuclear orientations as a method for determining hyperfine structure. Historical developments and future prospects for the technique are presented. The role that this technique can play in nuclear spectroscopy and the study of nuclei far from the beta stability are outlined

  9. WMI2, the Student's On-Line Symbolic Calculator

    Science.gov (United States)

    Kovacs, Zoltan

    2011-01-01

    Student activities focused on discovering mathematics play an important role in the teaching and learning process. WebMathematics Interactive (WMI2) was developed to offer a fast and user-friendly on-line web interface to enhance the quality of both theoretical and applied mathematics courses. For the teacher, in the classroom, it provides…

  10. ADAPTIVE CONTEXT PROCESSING IN ON-LINE HANDWRITTEN CHARACTER RECOGNITION

    NARCIS (Netherlands)

    Iwayama, N.; Ishigaki, K.

    2004-01-01

    We propose a new approach to context processing in on-line handwritten character recognition (OLCR). Based on the observation that writers often repeat the strings that they input, we take the approach of adaptive context processing. (ACP). In ACP, the strings input by a writer are automatically

  11. Microcomputers as on-line catalogs in special libraries.

    Science.gov (United States)

    Faust, J B

    1986-01-01

    This article discusses the rationale for the conversion of a card catalog to an on-line system in a special library owning approximately 4000 titles. Equipment, software, and procedures are described. Pros and cons of the use of a microcomputer for such a project, as well as costs and personnel needs, are outlined.

  12. Efficient and secure comparison for on-line auctions

    NARCIS (Netherlands)

    Damgard, Ivan; Geisler, M.; Kroigaard, M.; Pieprzyk, J.; Ghodosi, H.; Dawson, E.

    2007-01-01

    We propose a protocol for secure comparison of integers based on homomorphic encryption. We also propose a homomorphic encryption scheme that can be used in our protocol and makes it more efficient than previous solutions. Our protocol is well-suited for application in on-line auctions, both with

  13. The on-line asymmetric traveling salesman problem

    NARCIS (Netherlands)

    Ausiello, G.; Bonifaci, V.; Laura, L.

    2008-01-01

    We consider two on-line versions of the asymmetric traveling salesman problem with triangle inequality. For the homing version, in which the salesman is required to return in the city where it started from, we give a -competitive algorithm and prove that this is best possible. For the nomadic

  14. Facing regulatory challenges of on-line hemodiafiltration.

    Science.gov (United States)

    Kümmerle, Wolfgang

    2011-01-01

    On-line hemodiafiltration (on-line HDF) is the result of a vision that triggered multifarious changes in very different areas. Driven by the idea to offer better medical treatment for renal patients, technological innovations were developed and established that also constituted new challenges in the field of regulatory affairs. The existing regulations predominantly addressed the quality and safety of those products needed to perform dialysis treatment which were supplied by industrial manufacturers. However, the complexity of treatment system required for the provision of on-line fluids demanded a holistic approach encompassing all components involved. Hence, focus was placed not only on single products, but much more on their interfacing, and the clinical infrastructure, in particular, had to undergo substantial changes. The overall understanding of the interaction between such factors, quite different in their nature, was crucial to overcome the arising regulatory obstacles. This essay describes the evolution of the on-line HDF procedure from the regulatory point of view. A simplified diagram demonstrates the path taken from the former regulatory understanding to the realization of necessary changes. That achievement was only possible through 'management of preview' and consequent promotion of technical and medical innovations as well as regulatory re-evaluations. Copyright © 2011 S. Karger AG, Basel.

  15. Project development and commercialization of on-line analysis systems

    International Nuclear Information System (INIS)

    Watt, J.S.

    1997-01-01

    A project team first in the Australian Atomic Energy Commission (AAEC) and since 1982 in CSIRO has developed many on-line analysis systems for the mineral and energy industries. The development of these projects has followed a common pattern of laboratory R and D, field trials, commercialisation and technology transfer. This successful pattern is illustrated using examples of the development of systems for the on-line analysis of mineral slurries, for determination of the ash content of coal on conveyors, and for determination of the flow rates of oil, water and gas in pipelines. The first two systems are licensed to Australian companies, Amdel Ltd and Mineral Control Instrumentation Ltd. Both systems are used by industry worldwide, and are the market leaders for radioisotope gauges in their application field. The third system, the multiphase flow meter, was licensed in 1997 to Kvaerner FSSL Ltd of Aberdeen. This meter has even greater potential than the other two systems for economic benefit from its used and for numbers of installations. The on-line analysis systems have been developed to increase the productivity of the Australian mineral and energy industries, and to provide economic benefit to Australia. The economic benefit sought is predominantly improved process control based on use of the instrument, rather than from its sale. Sales of instruments are significant, however, with about A$80 million from the analysis systems and their derivatives since the 1970s. Some of the issues associated with the development of the on-line analysis system are outlined

  16. IOOS Data Portals and Uniform On-line Browse Capabilities

    Science.gov (United States)

    Howard, M.; Currier, R. D.; Kobara, S.; Gayanilo, F.

    2015-12-01

    The Gulf of Mexico Coastal Ocean Observing System Regional Association (GCOOS-RA) is one of eleven Regional Associations organized under the NOAA-led U.S. Integrated Ocean Observing System (IOOS) Program Office. Each of the RAs operate standards-based regional data portals designed to aggregate near real-time and historical observed data and modeled outputs from distributed providers and to offer these and derived products in standardized ways to a diverse set of users. The RA's portals are based on the IOOS Data and Communications Plan which describes the functional elements needed for an interoperable system. One of these elements is called "Uniform On-line Browse" which is an informational service designed primarily to visualize the inventory of a portal. An on-line browse service supports the end user's need to discover what parameters are available, to learn the spatial and temporal extend of the holdings, and to examine the character of the data (e.g, variability, gappiness, etc). These pieces of information help the end user decide if the data are fit for his/her purpose and to construct valid data requests. Note that on-line browse is a distinctly different activity than data analysis because it seeks to yield knowledge about the inventory and not about what the data mean. "Uniform" on-line browse is a service that takes advantage of the standardization of the data portal's data access points. Most portals represent station locations on a map. This is a view of the data inventory but these plots are rarely generated by pulling data through the standards-based services offered to the end users but through methods only available to the portal programmers. This work will present results of Uniform On-line browse tools developed within GCOOS-RA and their applicability to other RA portals.

  17. Mentoring Narratives ON-LINE:Teaching the Principalship

    Directory of Open Access Journals (Sweden)

    Allison I. Griffith

    2002-05-01

    Full Text Available The need to develop new models for preparation of school administrators has been a prominent concern in educational discourse in the last decade. Having been criticized for the inadequate preparation of the school leadership cadre, academic departments responsible for training future school administrators have had to revisit their approaches and to reframe their teaching philosophies to ensure the readiness of their graduates for the challenges and complexities of school leadership. This article reports on the new model of principals' training that has been used in York University's Principals' Qualification Program (PQP from the late 1990s onward. One component of the program brings traditional case methodology into a computer-mediated/on-line environment. The on-line cases are narratives from the everyday lives of the Ontario school administrators who serve as mentors in the on-line environment. Situating our discussion within the context of the rapidly changing educational landscape of Ontario, we focus on the PQP model to explore experientially generated case narratives as one method for teaching and learning the work of the local school administrator. We focus particularly on the teaching and learning embedded in computer-mediated or on-line case narratives used in training teachers for school leadership. We argue that the complexities of school leadership—the social, cultural, relational, ethical and moral context of school leadership—can be taught effectively through the reflective processes of on-line case narratives. We seek to contribute to the ongoing dialogue on the potential of new pedagogies and new technologies to help prepare the competent and responsible leaders for tomorrow's schools.

  18. Zinc injection in German PWR plants

    International Nuclear Information System (INIS)

    Streit, K.

    2004-01-01

    Operating experience acquired at PWR NNPs shows that zinc injection at low concentrations of 5 ppb is a very effective source term reduction measure. This method does not lead to any operating restrictions or other negative effects on plant systems and components. The nuclear industry has been very successful in reducing radiation exposures within the past two decades. Annual exposures could be significantly decreased and are now at a level of around 1 man-Sv per plant and year. This great success can mainly be attributed to the general commitment of plant operators to maintaining radiation exposures of workers in the controlled access area as low as reasonably achievable (ALARA principle). The ALARA principle, of course, also implies evaluation of the economic benefit of radiation protection measures. Radiation source term reduction has drawn increasing attention of plant operators in recent years. For the new PWRs cobalt-based alloys in the primary system have successively been eliminated already at the design and construction phase within the last decade. Use of wear-resistant cobalt-free substitute materials in combination with the general use of advanced alloys for the steam generator tubing of PWRs resulted in low values for the two most common sources of plant radiation fields, namely 58 Co and 60 Co. Investigations showed that the beneficial effect of zinc can be related to its high affinity for mixed spinel oxide phases, resulting in the following two basic effects: -Zinc is incorporated preferentially into the oxide layer on primary system surfaces and thus reduces pickup of 58 Co and 60 Co and - Zinc can displace cobalt isotopes from existing oxide layers. In German PWRs with Incoloy 800 steam generator tubing material (Ni-content -32%), the observed reductions correspond to a decrease in dose rates of around 10 to 15% per year and thus follow, as predicted, the half-life time of 60 Co. Overall reductions in high radiation areas are now in the range of

  19. On-line analysis of water contamination by organic compounds; On-line-Analytik der Wasserverschmutzung durch organische Substanzen

    Energy Technology Data Exchange (ETDEWEB)

    Wagt, R. van der; Vos, F. de [Skalar Analytical (Netherlands); Babichenko, S.; Poryvkina, L. [Institute of Ecology, Tallinn (Estonia)

    1999-08-01

    In many environmental applications decomposing the mixture of substances in the water into its various chemical ingredients, for subsequent analysis, is a very complicated task. The most productive approach to on-line diagnosis is to treat the object as an integral spectroscopic sample, characterized by certain specific Spectral Fluorescent Signatures (SFS). The SFS are recorded as a matrix of fluorescent intensity of organic compounds in water, in co-ordinates of excitation and emission spectra, providing a three-dimensional spectrum. Spectral windows of SFS are defined by fluorescent characteristics of basic groups of organic substances in the water sample. The novel Skalar Fluo Imager, based on this principle, is intended for the analysis of organic compounds in natural, domestic, and technological waters in an on-line mode. (orig.) [German] In vielen Umweltschutzanwendungen stellt die Auftrennung eines Substanzgemisches zum Zweck der Analyse eine sehr komplizierte Aufgabe dar. Ein erfolgversprechender Ansatz fuer eine on-line-Diagnostik besteht darin, das Objekt als integrale Spektroskopieprobe zu betrachten, die durch bestimmte spezifische Spektral-Fluoreszenz-Signaturen (SFS) charakterisiert wird. Diese werden als Fluoreszenz-Intensitaets-Matrix organischer Verbindungen in Wasser dargestellt, mit Anregungs- und Emissionsspektren als weiteren Koordinaten, wodurch ein dreidimensionales Spektrum entsteht. Spektrale Fenster der SFS sind definiert als Fluoreszenzcharakteristika von Funktionsgruppen organischer Substanzen in der Wasserprobe. Der auf diesem Prinzip basierende Skalar Fluo Imager ist fuer die on-line-Analyse organischer Bestandteile in natuerlichen, Haus- und technischen Waessern gedacht. (orig.)

  20. Automatic Delineation of On-Line Head-And-Neck Computed Tomography Images: Toward On-Line Adaptive Radiotherapy

    International Nuclear Information System (INIS)

    Zhang Tiezhi; Chi Yuwei; Meldolesi, Elisa; Yan Di

    2007-01-01

    Purpose: To develop and validate a fully automatic region-of-interest (ROI) delineation method for on-line adaptive radiotherapy. Methods and Materials: On-line adaptive radiotherapy requires a robust and automatic image segmentation method to delineate ROIs in on-line volumetric images. We have implemented an atlas-based image segmentation method to automatically delineate ROIs of head-and-neck helical computed tomography images. A total of 32 daily computed tomography images from 7 head-and-neck patients were delineated using this automatic image segmentation method. Manually drawn contours on the daily images were used as references in the evaluation of automatically delineated ROIs. Two methods were used in quantitative validation: (1) the dice similarity coefficient index, which indicates the overlapping ratio between the manually and automatically delineated ROIs; and (2) the distance transformation, which yields the distances between the manually and automatically delineated ROI surfaces. Results: Automatic segmentation showed agreement with manual contouring. For most ROIs, the dice similarity coefficient indexes were approximately 0.8. Similarly, the distance transformation evaluation results showed that the distances between the manually and automatically delineated ROI surfaces were mostly within 3 mm. The distances between two surfaces had a mean of 1 mm and standard deviation of <2 mm in most ROIs. Conclusion: With atlas-based image segmentation, it is feasible to automatically delineate ROIs on the head-and-neck helical computed tomography images in on-line adaptive treatments

  1. Radiation monitoring at Sizewell B PWR

    International Nuclear Information System (INIS)

    Hills, O.C.

    1992-01-01

    Radiation monitoring in Sizewell-B Power Station most significantly differs from that in existing UK Power Stations in two respects: firstly in the large number of on-line radiation monitors and secondly in the way that the monitors are linked into a fully-integrated, centralised data acquisition and display system, which can be accessed and viewed by the operators. An overview is given of how full process data is transmitted along data links between the Auxiliary Shut-Down Room, Technical Support Centre and Main Control Room, enabling Health Physics and other staff to access information from any radiation monitor. The permanently installed monitors together with the safety category, type and location are listed. As part of the Sizewell-B Process Plant control and instrumentation contract, NEI is to supply the Health Physics Instrumentation (HPI) and Process and Effluent Activity Monitoring System (PEAMS) (excluding the Primary Protection System) plus the Nuclear Sampling System (NSS). This paper concentrates on the HPI, and parts of the PEAMS and NSS for which NEI have the responsibility for system design, detail design, manufacture, site installation and commissioning. Section 2 briefly describes the sources of radiation at Sizewell-B; Sections 3, 4 and 5 describe the PEAMS, HPI and NSS respectively. Section 6 details the design of two of the Sizewell-B PEAMS subsystems. (Author)

  2. On-line soft sensing in upstream bioprocessing.

    Science.gov (United States)

    Randek, Judit; Mandenius, Carl-Fredrik

    2018-02-01

    This review provides an overview and a critical discussion of novel possibilities of applying soft sensors for on-line monitoring and control of industrial bioprocesses. Focus is on bio-product formation in the upstream process but also the integration with other parts of the process is addressed. The term soft sensor is used for the combination of analytical hardware data (from sensors, analytical devices, instruments and actuators) with mathematical models that create new real-time information about the process. In particular, the review assesses these possibilities from an industrial perspective, including sensor performance, information value and production economy. The capabilities of existing analytical on-line techniques are scrutinized in view of their usefulness in soft sensor setups and in relation to typical needs in bioprocessing in general. The review concludes with specific recommendations for further development of soft sensors for the monitoring and control of upstream bioprocessing.

  3. On-line corrosion monitoring in district heating systems

    DEFF Research Database (Denmark)

    Richter, Sonja; Thorarinsdottir, R.I.; Hilbert, Lisbeth Rischel

    2004-01-01

    ), Electrochemical Noise (EN) and Zero Resistance Ammetry (ZRA). Electrochemical Resistance (ER) has also been used to measure corrosion. The method traditionally only measures corrosion off-line but with newly developed high-sensitive ER technique developed by MetriCorr in Denmark, on-line monitoring is possible...... complicates the chemistry of the environment. Hydrogen sulphide is present in geothermal systems and can be formed as a by-product of sulphate-reducing-bacteria (SRB). The application of electrochemical methods makes on-line monitoring possible. These methods include: Linear Polarization Resistance (LPR....... In order to assess both general corrosion and localized corrosion, it is necessary to apply more than one monitoring technique simultaneously, ZRA or EN for measuring localized corrosion and LPR or ER for measuring general corrosion rate. The advantage of monitoring localized corrosion is indisputable...

  4. On-line sources of toxicological information in Canada

    International Nuclear Information System (INIS)

    Racz, William J.; Ecobichon, Donald J.; Baril, Marc

    2003-01-01

    This paper will provide an overview of the on-line resources available in toxicology in Canada. It will describe a brief history of The Society of Toxicology of Canada, with reference to other societies and also provide information on education, research and other resources related to toxicology. Toxicology in Canada emerged as a distinct and vibrant discipline following the thalidomide tragedy of the 1960s. In the pharmaceutical industry and government, toxicology was readily established as an essential component of drug development and safety, and as the need for toxicologists expanded, training programs were established, usually in collaboration with departments of pharmacology. In the last two to three decades other disciplines, environmental biology, analytical chemistry and epidemiology joined the ranks of toxicology. The on-line sources of toxicology information are rapidly expanding. This article describes those sources considered by the authors to be important from a national and international perspective. The majority of these sources are professional organizations and government agencies

  5. On-line chemistry monitoring for the secondary side

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Babcock and Wilcox (B and W) has developed a computerized water chemistry data acquisition and management system for nuclear plant secondary coolant systems. The Integrated Water Chemistry Monitoring System (IWCMS) provides on-line monitoring of conditions and rapid trend analysis of sampled data. So far it has been installed at GPU Three Mile Island unit 1 and at Toledo Edison Davis-Besse. The IWCMS meets the following utility needs for monitoring power plant chemistry: control of chemistry conditions to minimize corrosion and extend component/system life; continuous analysis of data from on-line detectors and grab samples; expediting of transient recovery actions with trend, alarm and evaluation capability; provision for rapid sharing of useful operational chemistry information; concentration of attention on evaluation instead of data manipulation. The system is composed of three functional parts: data acquisition hardware; PC-based computer system and customised system software. (author)

  6. On-line application of the PANTHER advanced nodal code

    International Nuclear Information System (INIS)

    Hutt, P.K.; Knight, M.P.

    1992-01-01

    Over the last few years, Nuclear Electric has developed an integrated core performance code package for both light water reactors (LWRs) and advanced gas-cooled reactors (AGRs) that can perform a comprehensive range of calculations for fuel cycle design, safety analysis, and on-line operational support for such plants. The package consists of the following codes: WIMS for lattice physics, PANTHER whole reactor nodal flux and AGR thermal hydraulics, VIPRE for LWR thermal hydraulics, and ENIGMA for fuel performance. These codes are integrated within a UNIX-based interactive system called the Reactor Physics Workbench (RPW), which provides an interactive graphic user interface and quality assurance records/data management. The RPW can also control calculational sequences and data flows. The package has been designed to run both off-line and on-line accessing plant data through the RPW

  7. UniFlex - Collaborative on-line learning environment tool

    Directory of Open Access Journals (Sweden)

    Ole Borch

    2004-05-01

    Full Text Available

    Første gang publiceret i UNEV nr. 2: E-læringsplatforme - muligheder og potentialer, januar - marts 2004, red. Tom Nyvand og Michael Pedersen. ISSN 1603-5518.

    Increasing demands for remote on-line education are changing the way teaching and learning is performed. New behavior in using pedagogy and supporting technology is needed to drive the learning process. To facilitate the use of services for selected activities to participants in distance education, a web site named UniFlex (University Flexible learning has been developed and brought into use. The site is a comprehensive set of bookmarks including course taking, upload/download, and - of special significance - collaborative on-line project work. UniFlex has been developed to meet the requirement for a simple and cheap personalized interactive site, supporting problem oriented and project organized study form, which has characterized Aalborg University for more than 27 years.

  8. UniFlex - Collaborative on-line learning environment tool

    Directory of Open Access Journals (Sweden)

    Ole Borch

    2004-03-01

    Full Text Available Første gang publiceret i UNEV nr. 2: E-læringsplatforme - muligheder og potentialer, januar - marts 2004, red. Tom Nyvand og Michael Pedersen. ISSN 1603-5518. Increasing demands for remote on-line education are changing the way teaching and learning is performed. New behavior in using pedagogy and supporting technology is needed to drive the learning process. To facilitate the use of services for selected activities to participants in distance education, a web site named UniFlex (University Flexible learning has been developed and brought into use. The site is a comprehensive set of bookmarks including course taking, upload/download, and - of special significance - collaborative on-line project work. UniFlex has been developed to meet the requirement for a simple and cheap personalized interactive site, supporting problem oriented and project organized study form, which has characterized Aalborg University for more than 27 years.

  9. On-line analyzers to distributed control system linking

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S.F.; Buchanan, B.R.; Sanders, M.A.

    1990-01-01

    The Analytical Development Section (ADS) of the Savannah River Laboratory is developing on-line analyzers to monitor various site processes. Data from some of the on-line analyzers (OLA's) will be used for process control by distributed control systems (DCS's) such as the Fisher PRoVOX. A problem in the past has been an efficient and cost effective way to get analyzer data onto the DCS data highway. ADS is developing a system to accomplish the linking of OLA's to PRoVOX DCS's. The system will be described, and results of operation in a research and development environment given. Plans for the installation in the production environment will be discussed.

  10. On-line Ramsey Numbers for Paths and Stars

    Directory of Open Access Journals (Sweden)

    Jaroslaw Grytczuk

    2008-08-01

    Full Text Available We study on-line version of size-Ramsey numbers of graphs defined via a game played between Builder and Painter: in one round Builder joins two vertices by an edge and Painter paints it red or blue. The goal of Builder is to force Painter to create a monochromatic copy of a fixed graph H in as few rounds as possible. The minimum number of rounds (assuming both players play perfectly is the on-line Ramsey number r(H of the graph H. We determine exact values of r(H for a few short paths and obtain a general upper bound r(Pn ≤ 4n-7. We also study asymmetric version of this parameter when one of the target graphs is a star Sn with n edges. We prove that r(Sn,H≤n ·e(H when H is any tree, cycle or clique.

  11. Laser systems for on-line laser ion sources

    International Nuclear Information System (INIS)

    Geppert, Christopher

    2008-01-01

    Since its initiation in the middle of the 1980s, the resonant ionization laser ion source has been established as a reliable and efficient on-line ion source for radioactive ion beams. In comparison to other on-line ion sources it comprises the advantages of high versatility for the elements to be ionized and of high selectivity and purity for the ion beam generated by resonant laser radiation. Dye laser systems have been the predominant and pioneering working horses for laser ion source applications up to recently, but the development of all-solid-state titanium:sapphire laser systems has nowadays initiated a significant evolution within this field. In this paper an overview of the ongoing developments will be given, which have contributed to the establishment of a number of new laser ion source facilities worldwide during the last five years.

  12. On-Line Metrology with Conoscopic Holography: Beyond Triangulation

    Directory of Open Access Journals (Sweden)

    Ignacio Álvarez

    2009-09-01

    Full Text Available On-line non-contact surface inspection with high precision is still an open problem. Laser triangulation techniques are the most common solution for this kind of systems, but there exist fundamental limitations to their applicability when high precisions, long standoffs or large apertures are needed, and when there are difficult operating conditions. Other methods are, in general, not applicable in hostile environments or inadequate for on-line measurement. In this paper we review the latest research in Conoscopic Holography, an interferometric technique that has been applied successfully in this kind of applications, ranging from submicrometric roughness measurements, to long standoff sensors for surface defect detection in steel at high temperatures.

  13. PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

    International Nuclear Information System (INIS)

    Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

    1988-01-01

    1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

  14. The Bochum on-line data acquisition system

    International Nuclear Information System (INIS)

    Paul, H.J.; Freiesleben, H.

    1986-01-01

    We describe an on-line data acquisition system based on a PDP 11 computer with CAMAC hardware. The software fully exploits the real-time features of the RSX-11M operating system. The basic characteristics of the program package, mainly written in FORTRAN 77, are: multitasking, shared common blocks, dynamical access to CAMAC hardware and data, and command orientated user interface. The system is particularly tailored for data acquisition in list mode of up to 64 parameters. (orig.)

  15. Recovery Strategies in On-Line Service Failure

    OpenAIRE

    Ozuem, Wilson; Lancaster, Geoff

    2013-01-01

    Despite a proliferation of a number of studies on service failures and recovery in e-service settings, there is a paucity of knowledge of ways in which service failures and recovery practices are implemented in the fashion industry. Drawing on constructivist perspective, this study offers a new perspective on an effective relational mechanism that would bridge the rupture between consumers and companies particularly in the on-line fashion sector. The analysis adds to studies on service failur...

  16. Understanding Cognitive Load Using On-line Dictionaries

    OpenAIRE

    Robert F. , Dilenschneider

    2017-01-01

    Cognitive Load Theory may useful for language instructors to understand how the look up conditions ofusing an on-line dictionary might influence learning. This paper first reviews previous studies that haveinvestigated dictionary use for vocabulary acquisition and reading comprehension Second, it explainsthe various elements of Cognitive Load Theory. Third, it describes how Cognitive Load Theory appliesto language learners' to learn unknown words and comprehend texts Last, it discusses the pe...

  17. On-line diagnostics for a real time system

    International Nuclear Information System (INIS)

    Sreenivasan, P.

    1976-01-01

    The purpose of an on-line diagnostics is to infuse the ability of self diagnosing in an online computer to enhance its dependability in a real time system. Such a diagnostics evolved for the CDPS of the Fast Breeder Test Reactor at Kalpakkam is reported. The two phases of the diagnostics, i.e., the malfunction detection and post detection action are described in some detail. (A.K.)

  18. Isocele I, the Orsay synchrocyclotron on-line separator

    International Nuclear Information System (INIS)

    Caruette, A.; Ferro, A.; Foucher, R.

    1976-01-01

    The main characteristics of the isotope separator Isocele 1 are described. This medium current separator was on line with the Orsay synchrocyclotron (155 MeV p, or 210 MeV 3 He) from March 1974 up to May 1975. Results obtained with different targets (Au, Bi, Er, Pt, Sn, Th) are summarized. They confirm the efficiency of medium current separators of this type [fr

  19. A decision support system for on-line leakage localization

    OpenAIRE

    Meseguer, Jordi; Mirats-Tur, Josep M.; Cembrano, Gabriela; Puig, Vicenç; Quevedo, Joseba; Pérez, Ramon; Sanz, Gerard; Ibarra, David

    2014-01-01

    This paper describes a model-driven decision-support system (software tool) implementing a model-based methodology for on-line leakage detection and localization which is useful for a large class of water distribution networks. Since these methods present a certain degree of complexity which limits their use to experts, the proposed software tool focuses on the integration of a method emphasizing its use by water network managers as a decision support system. The proposed software tool integr...

  20. On-line learning in radial basis functions networks

    OpenAIRE

    Freeman, Jason; Saad, David

    1997-01-01

    An analytic investigation of the average case learning and generalization properties of Radial Basis Function Networks (RBFs) is presented, utilising on-line gradient descent as the learning rule. The analytic method employed allows both the calculation of generalization error and the examination of the internal dynamics of the network. The generalization error and internal dynamics are then used to examine the role of the learning rate and the specialization of the hidden units, which gives ...

  1. On-line production of [11C]cyanogen bromide

    International Nuclear Information System (INIS)

    Westerberg, G.; Laangstroem, B.

    1997-01-01

    The electrophilic labelling precursor [ 11 C]cyanogen bromide was produced in 95% radiochemical yield (decay-corrected) from hydrogen [ 11 C]cyanide within 3 min from the end of bombardment using a simple and convenient solid-phase on-line procedure. The [ 11 C]cyanogen bromide has been used in the synthesis of a number of labelled compounds for use in positron emission tomography. (author)

  2. On-line Monitoring System for Power Transformers

    Directory of Open Access Journals (Sweden)

    Alexandru HOTEA

    2016-12-01

    Full Text Available Power transformers are the most important and expensive equipment from the electricity transmission system, so it is very important to know the real state of health of such equipment in every moment. De-energizing the power transformer accidentally due to internal defects can generate high costs. Annual maintenance proved to be ineffective in many cases to determine the internal condition of the equipment degradation due to faults rapidly evolving. An On-line Monitoring System for Power Transformers help real-time condition assessment and to detect errors early enough to take action to eliminate or minimize them. After abnormality detected, it is still important to perform full diagnostic tests to determine the exact condition of the equipment. On-line monitoring systems can help increase the level of availability and reliability of power transformers and lower costs of accidental interruption. This paper presents cases studies on several power transformers equipped with on-line monitoring systems from Transelectrica substation.

  3. Characterizing chemical systems with on-line computers and graphics

    International Nuclear Information System (INIS)

    Frazer, J.W.; Rigdon, L.P.; Brand, H.R.; Pomernacki, C.L.

    1979-01-01

    Incorporating computers and graphics on-line to chemical experiments and processes opens up new opportunities for the study and control of complex systems. Systems having many variables can be characterized even when the variable interactions are nonlinear, and the system cannot a priori be represented by numerical methods and models. That is, large sets of accurate data can be rapidly acquired, then modeling and graphic techniques can be used to obtain partial interpretation plus design of further experimentation. The experimenter can thus comparatively quickly iterate between experimentation and modeling to obtain a final solution. We have designed and characterized a versatile computer-controlled apparatus for chemical research, which incorporates on-line instrumentation and graphics. It can be used to determine the mechanism of enzyme-induced reactions or to optimize analytical methods. The apparatus can also be operated as a pilot plant to design control strategies. On-line graphics were used to display conventional plots used by biochemists and three-dimensional response-surface plots

  4. On-Line Fringe Tracking and Prediction at IOTA

    Science.gov (United States)

    Wilson, Edward; Mah, Robert; Lau, Sonie (Technical Monitor)

    1999-01-01

    The Infrared/Optical Telescope Array (IOTA) is a multi-aperture Michelson interferometer located on Mt. Hopkins near Tucson, Arizona. To enable viewing of fainter targets, an on-line fringe tracking system is presently under development at NASA Ames Research Center. The system has been developed off-line using actual data from IOTA, and is presently undergoing on-line implementation at IOTA. The system has two parts: (1) a fringe tracking system that identifies the center of a fringe packet by fitting a parametric model to the data; and (2) a fringe packet motion prediction system that uses characteristics of past fringe packets to predict fringe packet motion. Combined, this information will be used to optimize on-line the scanning trajectory, resulting in improved visibility of faint targets. Fringe packet identification is highly accurate and robust (99% of the 4000 fringe packets were identified correctly, the remaining 1% were either out of the scan range or too noisy to be seen) and is performed in 30-90 milliseconds on a Pentium II-based computer. Fringe packet prediction, currently performed using an adaptive linear predictor, delivers a 10% improvement over the baseline of predicting no motion.

  5. Review of trigger and on-line processors at SLAC

    International Nuclear Information System (INIS)

    Lankford, A.J.

    1984-07-01

    The role of trigger and on-line processors in reducing data rates to manageable proportions in e + e - physics experiments is defined not by high physics or background rates, but by the large event sizes of the general-purpose detectors employed. The rate of e + e - annihilation is low, and backgrounds are not high; yet the number of physics processes which can be studied is vast and varied. This paper begins by briefly describing the role of trigger processors in the e + e - context. The usual flow of the trigger decision process is illustrated with selected examples of SLAC trigger processing. The features are mentioned of triggering at the SLC and the trigger processing plans of the two SLC detectors: The Mark II and the SLD. The most common on-line processors at SLAC, the BADC, the SLAC Scanner Processor, the SLAC FASTBUS Controller, and the VAX CAMAC Channel, are discussed. Uses of the 168/E, 3081/E, and FASTBUS VAX processors are mentioned. The manner in which these processors are interfaced and the function they serve on line is described. Finally, the accelerator control system for the SLC is outlined. This paper is a survey in nature, and hence, relies heavily upon references to previous publications for detailed description of work mentioned here. 27 references, 9 figures, 1 table

  6. Designing effective on-line continuing medical education.

    Science.gov (United States)

    Zimitat, Craig

    2001-03-01

    The Internet, and new information and communication technologies available through the Internet, provides medical educators with an opportunity to develop unique on-line learning environments with real potential to improve physicians' knowledge and effect change in their clinical practice. There are approximately 100 websites offering on-line CME courses in the USA alone. However, few of these CME courses appear to be based on sound educational principles or CME research and may have little chance of achieving the broader goals of CME. The majority of these courses closely resemble their traditional counterparts (e.g. paper-based books are now electronic books) and appear to be mere substitutions for old-technology CME resources. Whilst some CME providers add unique features of the Internet to enrich their websites, they do not employ strategies to optimize the learning opportunities afforded by this new technology. The adoption of adult learning principles, reflective practice and problem-based approaches can be used as a foundation for sound CME course design. In addition, knowledge of Internet technology and the learning opportunities it affords, together with strategies to maintain participation and new assessment paradigms, are all needed for developing online CME. We argue for an evidence-based and strategic approach to the development of on-line CME courses designed to enhance physician learning and facilitate change in clinical behaviour.

  7. On-line radiation teaching materials using IT technology

    International Nuclear Information System (INIS)

    Inoue, Hiroyoshi

    2005-01-01

    We developed the on-line radiation teaching materials using the Internet, in order to provide the teaching support materials of atomic power and radiation educations in on-school study, as well as to create the complementary study system in off-school study. The themes of teaching materials were selected from requests by teachers. In the case of an elementary school, the teaching material 'an environmental problem and atomic power' was created as the aggregate of each content for study without boundary between subjects. The teaching material 'medical treatment and radiation' was created for junior high school students to raise the individual knowledge. In the case of a high school, the teaching material nucleus and radiation' was prepared to supplement the physical study of students. The on-line teaching materials were tried to 300 junior high school and high school students, 68% of students answered that the teaching material is effective to understand atomic power and radiation, though 17% answered they were not effective. Although there are problems to prepare IT learning equipments and learning follow-up system in the material, it is suggested that the on-line teaching materials will provide the novel learning system including debates for the study. This method has no limitation of time and place. (author)

  8. On-line fouling monitor for heat exchangers

    International Nuclear Information System (INIS)

    Tsou, J.L.

    1995-01-01

    Biological and/or chemical fouling in utility service water system heat exchangers adversely affects operation and maintenance costs, and reduced heat transfer capability can force a power deaerating or even a plant shut down. In addition, service water heat exchanger performance is a safety issue for nuclear power plants, and the issue was highlighted by NRC in Generic Letter 89-13. Heat transfer losses due to fouling are difficult to measure and, usually, quantitative assessment of the impact of fouling is impossible. Plant operators typically measure inlet and outlet water temperatures and flow rates and then perform complex calculations for heat exchanger fouling resistance or ''cleanliness''. These direct estimates are often imprecise due to inadequate instrumentation. Electric Power Research Institute developed and patented an on-line condenser fouling monitor. This monitor may be installed in any location within the condenser; does not interfere with routine plant operations, including on-line mechanical and chemical treatment methods; and provides continuous, real-time readings of the heat transfer efficiency of the instrumented tube. This instrument can be modified to perform on-line monitoring of service water heat exchangers. This paper discusses the design, construction of the new monitor, and algorithm used to calculate service water heat exchanger fouling

  9. Denmark's on - line early warning radiation monitoring network

    International Nuclear Information System (INIS)

    Walmod-Larsen, O.; Lippert, J.

    1990-01-01

    In Denmark an emergency response coordination committee was set up to cope with the problems after the Chernobyl accident with participation of all relevant authorities. For help in an emergency situation the ARGOS (Accident Reporting and Guiding Operational System), system will be put into use. The ARGOS emergency evaluation computer system, which has been developed in cooperation with the Danish Environmental Protection Agency, is in operation in connection with the emergency planning for the east region of Denmark with regard to the Swedish nuclear power plant operating at Barseback. Inputs of measurement data are on-line available on data screens for evaluation in the emergency coordination centers, presented on suitable geographical maps, showing iso-contours calculated from the input. In case of an alert situation other systems can be put in operation, f.ex. mobile measuring units from the CDEPA's local, operational emergency centers. Their readings can then be put into the computing system parallel to the on-line stations and be presented by the ARGOS-system for evaluation in the emergency command centers. If another national authority in an alert situation requests a transfer of measurement data, and if this is agreed upon by the competent Danish authority, then the transfer can be arranged from the ARGOS-system, through agreed transmission channels. At present the ARGOS system is being improved and expanded by RNL to cover the whole Danish region and to present measurements from the on-line warning system

  10. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  11. On-line monitoring of boiling crevice chemistry evolution

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, C.B.; Oh, S.; Park, B.G.; Hwang, I.S. [Department of Nuclear Engineering, Seoul National Univ. (Korea, Republic of); Rhee, I.H. [Department of Chemical Engineering, Soonchunhyang Univ. (Korea, Republic of); Kim, U.C.; Na, J.W. [Korea Atomic Energy Research Inst., Daejon (Korea, Republic of)

    2002-07-01

    In a locally restricted geometry on the secondary side of steam generator (SG) in a pressurized water reactor (PWR), impurities in bulk water can be concentrated by boiling process to extreme pH that may then accelerate the corrosion of tubing and adjacent materials. To simulate a real SG tubesheet crevice, a high temperature/high pressure (HT/HP) crevice simulation system was constructed. Primary water was pumped at a high flow rate through a 3/4'' outer-diameter tubing and a crevice section was made on the outer diameter (OD) side of the tubing. The simulated crevice area was monitored with thermocouples and electrodes for the measurement of temperature and electrochemical corrosion potential (ECP), respectively, in the crevice as well as free span. A secondary solution composed of 50 ppm Na and 200 ppb hydrogen (H{sub 2}) was supplied at a flow rate of about 4 L/hr. In an open tubesheet crevice with 0.15 mm radial gap and 40 mm depth, axial distributions of temperature and ECP were measured as a function of time and available superheat. Sodium hydroxide (NaOH) concentration process in the crevice and the resultant evolution of crevice boiling regions were characterized from temperature and ECP data. Measured data for an open crevice showed a similar behavior to predictions by a thermodynamic equilibrium code. Magnetite-packed crevice had much longer time to reach a steady state than open crevice. (authors)

  12. Operating system design of parallel computer for on-line management of nuclear pressurised water reactor cores

    International Nuclear Information System (INIS)

    Gougam, F.

    1991-04-01

    This study is part of the PHAETON project which aims at increasing the knowledge of safety parameters of PWR core and reducing operating margins during the reactor cycle. The on-line system associates a simulator process to compute the three dimensional flux distribution and an acquisition process of reactor core parameters from the central instrumentation. The 3D flux calculation is the most time consuming. So, for cost and safety reasons, the PHAETON project proposes an approach which is to parallelize the 3D diffusion calculation and to use a computer based on parallel processor architecture. This paper presents the design of the operating system on which the application is executed. The routine interface proposed, includes the main operations necessary for programming a real time and parallel application. The primitives include: task management, data transfer, synchronisation by event signalling and by using the rendez-vous mechanisms. The primitives which are proposed use standard softwares like real-time kernel and UNIX operating system [fr

  13. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, Joao Claudio B.

    2015-01-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  14. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  15. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  16. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  17. Application of digital control in Japanese PWR Plants

    International Nuclear Information System (INIS)

    Taguchi, S.; Kondo, Y.; Teranishi, S.; Matsumiya, M.; Takashima, M.; Nagai, T.

    1986-01-01

    More reliable and flexible control system to improve the plant availability and operability is constantly demanded. In order to answer the demands, digital control systems are being applied to Japanese PWR plants. Microprocessor-based digital control systems are widely used in other industries and show good performance. The digital control system has been already applied to the chemical and volume control system and the radioactive waste disposal system in the operating plants. These systems have been working as expected and demonstrating good performances. The digital control system for the reactor control system, which is the main control system of the PWR plants, is being developed. The design of the system has been already finished and the verification/validation process is now in progress

  18. A concept of PWR using plate and shell heat exchangers

    International Nuclear Information System (INIS)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de

    2015-01-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  19. The development of flow test technology for PWR fuel assembly

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

    1988-05-01

    KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

  20. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    Matzie, R.A.; Daleas, R.S.; Miller, D.D.

    1980-11-01

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  1. Electrical and control aspects of the Sizewell B PWR

    International Nuclear Information System (INIS)

    1992-01-01

    The pressurized water reactor, Sizewell-B, which is being built in Suffolk is well on in its construction schedule. This conference looked at the electrical and control aspects of the first PWR to be built in the United Kingdom. Although based on the standard Westinghouse PWR design, modifications have been made to meet the particular requirements of the site and the UK licensing regulations. There are 11 papers on all aspects of the electrical systems, 5 papers on the cables and cable installation, 5 on the main control rooms and auxiliary shutdown room, 5 on the integrated system and centralised operation, 6 on the monitoring and protection systems and 9 on the reactor protection systems. All 41 are indexed separately. (UK)

  2. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  3. PWR accident management realated tests: some Bethsy results

    International Nuclear Information System (INIS)

    Clement, P.; Chataing, T.; Deruaz, R.

    1993-01-01

    The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

  4. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  5. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  6. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  7. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  8. Report on the PWR-radiation protection/ALARA Committee

    Energy Technology Data Exchange (ETDEWEB)

    Malone, D.J. [Consumers Power Co., Covert, MI (United States)

    1995-03-01

    In 1992, representatives from several utilities with operational Pressurized Water Reactors (PWR) formed the PWR-Radiation Protection/ALARA Committee. The mission of the Committee is to facilitate open communications between member utilities relative to radiation protection and ALARA issues such that cost effective dose reduction and radiation protection measures may be instituted. While industry deregulation appears inevitable and inter-utility competition is on the rise, Committee members are fully committed to sharing both positive and negative experiences for the benefit of the health and safety of the radiation worker. Committee meetings provide current operational experiences through members providing Plant status reports, and information relative to programmatic improvements through member presentations and topic specific workshops. The most recent Committee workshop was facilitated to provide members with defined experiences that provide cost effective ALARA performance.

  9. Assessment of PWR plutonium burners for nuclear energy centers

    International Nuclear Information System (INIS)

    Frankel, A.J.; Shapiro, N.L.

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible

  10. A concept of PWR using plate and shell heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir; Andrade, Delvonei Alves de, E-mail: luciano.ondir@gmail.com, E-mail: delvonei@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    In previous work it was verified the physical possibility of using plate and shell heat exchangers for steam generation in a PWR for merchant ships. This work studies the possibility of using GESMEX commercial of the shelf plate and shell heat exchanger of series XPS. It was found it is feasible for this type of heat exchanger to meet operational and accidental requirements for steam generation in PWR. Additionally, it is proposed an arrangement of such heat exchangers inside the reactor pressure vessel. Such arrangement may avoid ANSI/ANS51.1 nuclear class I requirements on those heat exchangers because they are contained in the reactor coolant pressure barrier and play no role in accidental scenarios. Additionally, those plates work under compression, preventing the risk of rupture. Being considered non-nuclear safety, having a modular architecture and working under compression may turn such architectural choice a must to meet safety objectives with improved economics. (author)

  11. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  12. Surveillance systems (PWR) - loose parts monitoring - vibration monitoring - leakage detection

    International Nuclear Information System (INIS)

    Schuette, A.; Blaesig, H.

    1982-01-01

    The contribution is engaged in the task and the results of the loose parts monitoring and the vibration monitoring following from the practice at the PWR of Biblis. First a description of both systems - location and type of the sensors used, the treatment of the measurements and the indications - is given. The results of the analysis of some events picked up by the surveillance systems are presented showing applicabilty and benefit of such systems. (orig.)

  13. A model to calculate the burn of gadolinium in PWR

    International Nuclear Information System (INIS)

    Sannazzaro, L.R.

    1983-01-01

    A cell model to calculate the burnup of a PWR fuel element with gadolinium as a poison, projected by KWU, is presented. With the model proposed, the burn of the gadolinium isotopes is analyzed, as well as the effect of these isotopes in the fuel element behaviour. The results obtained with this cell model are compared with those obtained by a conventional cell model. (E.G.) [pt

  14. Conversion rate for PWR reactors in thorium cycle

    International Nuclear Information System (INIS)

    Angelkorte, G.M.

    1980-01-01

    This work concerns to the determination of the conversion-rate for a PWR reactor with an enrichment of 7.47%, considering a cell, geometrically equal to Angra I, composed by Thorium and U-238 in a 1:1 relation. The study was performed considering neutrons of one and two groups of energy, according to the suggestion from other authors sup(1,2). It was also performed a study about the production and consumption of fissile material. (author)

  15. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  16. Measured performance of four PWR liquid radioactive waste treatment systems

    International Nuclear Information System (INIS)

    McIsaac, C.V.; Mandler, J.W.; Stalker, A.C.

    1980-01-01

    This paper presents results of a study of the liquid radwaste treatment and boron recovery systems of four operating PWR power plants. The performance of a given system was determined from measurements of radionuclide inventories in samples drawn from demineralizers, evaporators, filters, and gaseous cleanup systems. The plants at which measurements were made are Fort Calhoun, Zion 1 and 2, Turkey Point 3 and 4, and Rancho Seco

  17. Uranium savings on a once through PWR fuel cycle

    International Nuclear Information System (INIS)

    Cupo, J.V.

    1980-01-01

    A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

  18. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  19. Stress analysis on a PWR pressure vessel support structure

    International Nuclear Information System (INIS)

    Cruz, J.R.B.; Mattar Neto, M.; Jesus Miranda, C.A. de.

    1992-01-01

    The paper presents the stress analysis of a research PWR vessel support structure. Different geometries and thermal boundary conditions are evaluated. The finite element analysis is performed using ANSYS program. The ASME Section III criteria are applied for the stress verification and the following points are discussed: stress classification and linearization; jurisdictional boundary between ASME Subsection NB (Class 1 Components) and Subsection NF (Component Supports). (author)

  20. ORNL: PWR-BDHT analysis procedure, a preliminary overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The computer programs currently used in the analysis of the ORNL-PWR Blowdown Heat Transfer Separate-Effects Program are overviewed. The current linkages and relationships among the programs are given along with general comments about the future directions of some of these programs. The overview is strictly from the computer science point of view with only minimal information concerning the engineering aspects of the analysis procedure

  1. Improvement on main control room for Japanese PWR plants

    International Nuclear Information System (INIS)

    Matsumiya, Masayuki

    1996-01-01

    The main control room which is the information center of nuclear power plant has been continuously improved utilizing the state of the art ergonomics, a high performance computer, computer graphic technologies, etc. For the latest Japanese Pressurized Water Reactor (PWR) plant, the CRT monitoring system is applied as the major information source for facilitating operators' plant monitoring tasks. For an operating plant, enhancement of monitoring and logging functions has been made adopting a high performance computer

  2. A comparative study of fuel management in PWR reactors

    International Nuclear Information System (INIS)

    Barroso, D.E.G.; Nair, R.P.K.; Vellozo, S.O.

    1981-01-01

    A study about fuel management in PWR reactors, where not only the conventional uranium cycle is considered, but also the thorium cycle as an alternative is presented. The final results are presented in terms of U 3 O 8 demand and SWU and the approximate costs of the principal stages of the fuel cycle, comparing with the stardand cycle without recycling. (E.G.) [pt

  3. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.) [pt

  4. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

  5. Fire experiences: principal lessons learned, application in PWR power plants

    International Nuclear Information System (INIS)

    Schoemacker, M.

    1984-01-01

    The article reviews the principal design rules to be borne in mind for PWR nuclear units installation. These rule takes into account: the specific character of materials involved (safety aspect for nuclear construction), experience acquired as a result of fires in EDF production units, and the results obtained from tests carried out by the EDF at Fort de Chelles between 1980 and 1982, especially in the field of PVC cables [fr

  6. Natural circulation in a scaled PWR integral test facility

    International Nuclear Information System (INIS)

    Kiang, R.L.; Jeuck, P.R. III

    1987-01-01

    Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

  7. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  8. ORNL-PWR BDHT analysis procedure: an overview

    International Nuclear Information System (INIS)

    Cliff, S.B.

    1978-01-01

    The key computer programs currently used by the analysis procedure of the ORNL-PWR Blowdown Heat Transfer Separate Effects Program are overviewed with particular emphasis placed on their interrelationships. The major modeling and calculational programs, COBRA, ORINC, ORTCAL, PINSIM, and various versions of RELAP4, are summarized and placed into the perspective of the procedure. The supportive programs, REDPLT, ORCPLT, BDHTPLOT, OXREPT, and OTOCI, and their uses are described

  9. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  10. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  11. Acceptance test for 900 MWe PWR unit replacement steam generators

    International Nuclear Information System (INIS)

    Gourguechon, B.

    1993-01-01

    During the first half of 1994, the Gravelines 1 steam generators will be replaced (SG replacement procedure). The new SG's differ from the former components notably by the alloy used for the tube bundle, in this case, the high chromium content Inconel 690. So, from this standpoint, they are to be considered as PWR 900 replacement SG first models and their thermal efficiency has consequently to be assessed. This will provide an opportunity of ensuring that the performance of the components delivered is in compliance with requirements and of making the necessary provisions if significant deviations are observed. The EFMT branch, which has been in charge of the instrumentation and acceptance of the different SG first models since the first PWR plants were commissioned, will be responsible for the acceptance tests and the ultimate validation of a performance assessment procedure applicable to the future replacement steam generators. The methods and tests proposed for SG expert appraisal are based on consideration of the importance of primary measurement quality for satisfactory SG assessment and of the new test facilities with which the 900 and 1 300 PWR plants are gradually being equipped. These facilities provide an on-site computer environment for tests compatible with the tools (PATTERN, etc.) used at EFMT and in other departments. This test is the first of this kind performed by EFMT and the test facility of a nuclear power plant. (author). 6 figs

  12. Performance of high burned PWR fuel during transient

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio

    1992-01-01

    In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

  13. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  14. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  15. Actinides transmutation - a comparison of results for PWR benchmark

    International Nuclear Information System (INIS)

    Claro, Luiz H.

    2009-01-01

    The physical aspects involved in the Partitioning and Transmutation (P and T) of minor actinides (MA) and fission products (FP) generated by reactors PWR are of great interest in the nuclear industry. Besides these the reduction in the storage of radioactive wastes are related with the acceptability of the nuclear electric power. From the several concepts for partitioning and transmutation suggested in literature, one of them involves PWR reactors to burn the fuel containing plutonium and minor actinides reprocessed of UO 2 used in previous stages. In this work are presented the results of the calculations of a benchmark in P and T carried with WIMSD5B program using its new cross sections library generated from the ENDF-B-VII and the comparison with the results published in literature by other calculations. For comparison, was used the benchmark transmutation concept based in a typical PWR cell and the analyzed results were the k∞ and the atomic density of the isotopes Np-239, Pu-241, Pu-242 and Am-242m, as function of burnup considering discharge of 50 GWd/tHM. (author)

  16. A scheme of better utilization of PWR spent fuels

    International Nuclear Information System (INIS)

    Chung, Bum Jin; Kang, Chang Soon

    1991-01-01

    The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

  17. In the Jungle of Astronomical On--line Data Services

    Science.gov (United States)

    Egret, D.

    The author tried to survive in the jungle of astronomical on--line data services. In order to find efficient answers to common scientific data retrieval requests, he had to collect many pieces of information, in order to formulate typical user scenarios, and try them against a number of different data bases, catalogue services, or information systems. He discovered soon how frustrating treasure coffers may be when their keys are not available, but he realized also that nice widgets and gadgets are of no help when the information is not there. And, before long, he knew he would have to navigate through several systems because no one was yet offering a general answer to all his questions. I will present examples of common user scenarios and show how they were tested against a number of services. I will propose some elements of classification which should help the end-user to evaluate how adequate the different services may be for providing satisfying answers to specific queries. For that, many aspects of the user interaction will be considered: documentation, access, query formulation, functionalities, qualification of the data, overall efficiency, etc. I will also suggest possible improvements to the present situation: the first of them being to encourage system managers to increase collaboration between one another, for the benefit of the whole astronomical community. The subjective review I will present, is based on publicly available astronomical on--line services from the U.S. and from Europe, most of which (excepting the newcomers) were described in ``Databases and On-Line Data in Astronomy", (Albrecht & Egret, eds, 1991): this includes databases (such as NED and Simbad ), catalog services ( StarCat , DIRA , XCatScan , etc.), and information systems ( ADS and ESIS ).

  18. On-line implant reconstruction in HDR brachytherapy

    International Nuclear Information System (INIS)

    Kolkman-Deurloo, Inger-Karine K.; Kruijf, Wilhelmus J.M. de; Levendag, Peter C.

    2006-01-01

    Background and purpose: To evaluate the accuracy of on-line planning in an Integrated Brachytherapy Unit (IBU) using dedicated image distortion correction algorithms, correcting the geometric distortion and magnetic distortion separately, and to determine the effect of the reconstruction accuracy on clinical treatment plans in terms of deviations in treatment time and dose. Patients and methods: The reconstruction accuracy has been measured using 20 markers, positioned at well known locations in a QA phantom. Treatment plans of two phantoms representing clinical implant geometries, have been compared with reference plans to determine the effect of the reconstruction accuracy on the treatment plan. Before clinical introduction, treatment plans of three representative patients, based on on-line reconstruction, have been compared with reference plans. Results: The average reconstruction error for 10 in. images reduces from -0.6 mm (range -2.6 to +1.0 mm) to -0.2 mm (range -1.2 to +0.6 mm) after image distortion correction and for 15 in. images from 0.8 mm (range -0.5 to +3.0 mm) to 0.0 mm (range -0.8 to +0.8 mm). The error in case of eccentric positioning of the phantom, i.e. 0.8 mm (range -1.0 to +3.3 mm), reduces to 0.1 mm (range -0.5 to +0.9 mm). Correction of the image distortions reduces the deviation in the calculated treatment time of maximally 2.7% to less than 0.8% in case of eccentrically positioned clinical phantoms. The deviation in the treatment time or reference dose in the plans based on on-line reconstruction with image distortion correction of the three patient examples is smaller than 0.3%. Conclusions: Accurate on-line implant reconstruction using the IBU localiser and dedicated correction algorithms separating the geometric distortion and the magnetic distortion is possible. The results fulfill the minimum requirements as imposed by the Netherlands Commission on Radiation Dosimetry (NCS) without limitations regarding the usable range of the field

  19. On-line data processing for scintillation camera

    International Nuclear Information System (INIS)

    Ueyanagi, Hideo

    1974-01-01

    To process on-line the information from scintillation cameras, the computing/processing device of wired program type, with a specialized mini-computer for the processing, is generally used; and by this method, the data processing is done by the users of scintillation cameras. In the device with a mini-computer with processing programming by software, the almost all items in processing can be executed; but the operation requires some skill. With mini-computer operation, on the other hand, there are the problems of data-point number, process-time reduction, and image storage for both long and short terms. (Mori, K.)

  20. Robust Control Methods for On-Line Statistical Learning

    Directory of Open Access Journals (Sweden)

    Capobianco Enrico

    2001-01-01

    Full Text Available The issue of controlling that data processing in an experiment results not affected by the presence of outliers is relevant for statistical control and learning studies. Learning schemes should thus be tested for their capacity of handling outliers in the observed training set so to achieve reliable estimates with respect to the crucial bias and variance aspects. We describe possible ways of endowing neural networks with statistically robust properties by defining feasible error criteria. It is convenient to cast neural nets in state space representations and apply both Kalman filter and stochastic approximation procedures in order to suggest statistically robustified solutions for on-line learning.

  1. Development of an on-line radon monitoring system

    International Nuclear Information System (INIS)

    Guo Huiping; Shang Aiguo; Liu Junfeng; Zhou Chunlin; Di Yuming

    2004-01-01

    Of the actual demand by the strategic missile troops, the author has successfully developed a specially designed passive diffusion collecting chamber to collect the decay products of radon by high voltage static electricity, and using the single-chip microcomputer to reckon the radon concentration in air, which is actually a portable, continuous and automatic on-line monitoring instrument. It was made into a four-slot standard plug-in board of a NIM, and it functions as auto data memory, data process, display, over-threshold alarming and so on. (authors)

  2. Integrated on-line accelerator modeling at CEBAF

    International Nuclear Information System (INIS)

    Bowling, B.A.; Shoaee, H.; Van Zeijts, J.; Witherspoon, S.; Watson, W.

    1995-01-01

    An on-line accelerator modeling facility is currently under development at CEBAF. The model server, which is integrated with the EPICS control system, provides coupled and 2nd-order matrices for the entire accelerator, and forms the foundation for automated model- based control and diagnostic applications. Four types of machine models are provided, including design, golden or certified, live, and scratch or simulated model. Provisions are also made for the use of multiple lattice modeling programs such as DIMAD, PARMELA, and TLIE. Design and implementation details are discussed. 2 refs., 4 figs

  3. On-line Dynamic Security Assessment in Power Systems

    DEFF Research Database (Denmark)

    Weckesser, Johannes Tilman Gabriel

    and solar radiation. Moreover, ongoing research suggests that demand response will be introduced to maintain power balance between generation and consumption at all times. Due to these changes the operating point of the power system will be less predictable and today’s stability and security assessment...... for early prediction of critical voltage sags is described. The method’s performance is compared to other prediction approaches. The results show that the proposed method succeeds in early, accurately and consistently predicting critically low voltage sags. An efficient on-line DSA not only identifies...

  4. Enhancing Learner Autonomy in an On-line Editing Programme

    Directory of Open Access Journals (Sweden)

    Hebe Wong

    2011-09-01

    Full Text Available Little (1999 argues that in formal educational contexts, “the basis of learner autonomy is acceptance of responsibility for one’s own learning” (p.11. An autonomous learner takes responsibility for various aspects of learning (Benson & Voller, 1997; Holec, 1981. This study examines how learner autonomy opportunities were provided at various stages of writing in an on-line editing programme for a group of electronic engineering students and how the students took charge of their language learning when receiving feedback on their technical writing. The impact on their own learning effectiveness of the decisions students made is also discussed.

  5. Investigating on-line pornography at the University of Johannesburg

    Directory of Open Access Journals (Sweden)

    P. Laughton

    2007-12-01

    (61,6% aged between 19 years and 21 years. Of the respondents, 36,7% indicated that exposure to unsolicited pornography did not bother them. When asked to what extent students should have access to pornography, 60,5% stated 'None' while 32,6% believed that 'Restricted' access should be granted for research purposes and 6,9% believed that students should be granted 'Total' access to pornography. Results from the research will be used to manage access to on-line resources at the University of Johannesburg better.

  6. On-line core monitoring with CORE MASTER / PRESTO

    International Nuclear Information System (INIS)

    Lindahl, S.O.; Borresen, S.; Ovrum, S.

    1986-01-01

    Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

  7. Core on-line monitoring and computerized procedures systems

    International Nuclear Information System (INIS)

    Gangloff, W.C.

    1986-01-01

    The availability of operating nuclear power plants has been affected significantly by the difficulty people have in coping with the complexity of the plants and the operating procedures. Two ways to use modern computer technology to ease the burden of coping are discussed in this paper, an on-line core monitoring system with predictive capability and a computerized procedures system using live plant data. These systems reduce human errors by presenting information rather than simply data, using the computer to manipulate the data, but leaving the decisions to the plant operator

  8. Lower Bounds and Semi On-line Multiprocessor Scheduling

    Directory of Open Access Journals (Sweden)

    T.C. Edwin Cheng

    2003-10-01

    Full Text Available We are given a set of identical machines and a sequence of jobs from which we know the sum of the job weights in advance. The jobs have to be assigned on-line to one of the machines and the objective is to minimize the makespan. An algorithm with performance ratio 1.6 and a lower bound of 1.5 is presented. This improves recent results by Azar and Regev who published an algorithm with performance ratio 1.625 for the less general problem that the optimal makespan is known in advance.

  9. Adaptive calibration method with on-line growing complexity

    Directory of Open Access Journals (Sweden)

    Šika Z.

    2011-12-01

    Full Text Available This paper describes a modified variant of a kinematical calibration algorithm. In the beginning, a brief review of the calibration algorithm and its simple modification are described. As the described calibration modification uses some ideas used by the Lolimot algorithm, the algorithm is described and explained. Main topic of this paper is a description of a synthesis of the Lolimot-based calibration that leads to an adaptive algorithm with an on-line growing complexity. The paper contains a comparison of simple examples results and a discussion. A note about future research topics is also included.

  10. The development of an on-line gold analyser

    International Nuclear Information System (INIS)

    Robert, R.V.D.; Ormrod, G.T.W.

    1982-01-01

    An on-line analyser to monitor the gold in solutions from the carbon-in-pulp process is described. The automatic system is based on the delivery of filtered samples of the solutions to a distribution valve for measurement by flameless atomic-absorption spectrophotometry. The samples is introduced by the aerosol-deposition method. Operation of the analyser on a pilot plant and on a full-scale carbon-in-pulp plant has shown that the system is economically feasible and capable of providing a continuous indication of the efficiency of the extraction process

  11. NNDC [National Nuclear Data Center] on-line services documentation

    International Nuclear Information System (INIS)

    Dunford, C.L.; Burrows, T.W.; Tuli, J.K.

    1987-01-01

    This document summarizes and describes how to access the on-line services available from the National Nuclear Data Center (NNDC) located at Brookhaven National Laboratory. The services are available free of cost to US Department of Energy, its contractors and others who support the NNDC or supply data to the NNDC. Four of the center's data bases are now accessible to non-NNDC scientists via remote connection to the center's VAX 11/780. To use this service, you must have a terminal with access by either a telephone line or the PHYSNET network. A VT100 terminal or a terminal with VT-100 emulation is recommended but not required

  12. Efficient and Secure Comparison for On-Line Auctions

    DEFF Research Database (Denmark)

    Damgård, Ivan Bjerre; Krøigaard, Mikkel; Geisler, Martin Joakim

    2007-01-01

    We propose a protocol for secure comparison of integers based on homomorphic encryption. We also propose a homomorphic encryption scheme that can be used in our protocol and makes it more efficient than previous solutions. Our protocol is well-suited for application in on-line auctions, both...... with respect to functionality and performance. It minimizes the amount of information bidders need to send, and for comparison of 16 bit numbers with security based on 1024 bit RSA (executed by two parties), our implementation takes 0.28 seconds including all computation and communication. Using precomputation...

  13. On-line computing in a classified environment

    International Nuclear Information System (INIS)

    O'Callaghan, P.B.

    1982-01-01

    Westinghouse Hanford Company (WHC) recently developed a Department of Energy (DOE) approved real-time, on-line computer system to control nuclear material. The system simultaneously processes both classified and unclassified information. Implementation of this system required application of many security techniques. The system has a secure, but user friendly interface. Many software applications protect the integrity of the data base from malevolent or accidental errors. Programming practices ensure the integrity of the computer system software. The audit trail and the reports generation capability record user actions and status of the nuclear material inventory

  14. On-line system for investigation of atomic structure

    International Nuclear Information System (INIS)

    Amus'ya, M.Ya.; Chernysheva, L.V.

    1983-01-01

    A description of the on-line ATOM system is presented that enables to investigate the structure of atomic electron shells and their interactions with different scattering particles-electrons, positronse photons, mesons - with the use of computerized numerical solutions. The problem is stated along with mathematical description of atomic properties including theoretical and numerical models for each investigated physical process. The ATOM system structure is considered. The Hartree-Fock method is used to determine the wave functions of the ground and excited atomic states. The programs are written in the ALGOL langauge. Different atomic characteristics were possible to be calculated for the first time with an accuracy exceeding an experimental one

  15. ELT-MELAS analyzer and its on-line programs

    International Nuclear Information System (INIS)

    Anikeev, V.B.; Berezhnoj, V.A.; Glupova

    1976-01-01

    ELT-MELAS device constructed for an automatic analysis of pictures from big bubble chambers is described. It is controlled by a medium-size ICL-1903A computer and has two measuring modes: analysis of the ''agreement'' signal and digitation of slice-scans. Main features of the hardware and of on-line controlling and diagnostic software are presented. The test results of the MELAS complex as well as preliminary results of the scan-slice measurements of pictures from 15sup(') chamber are given

  16. Study on virtual simulation technology for operation and control of PWR

    International Nuclear Information System (INIS)

    Fang Baoguo; Zhang Dafa; Lin Yajun

    2006-01-01

    The way to build graphical models of PWR with MultiGen Creator is discussed, and the three-dimensional model used in the virtual simulation is built. The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer. And, all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR. (authors)

  17. Overview of the Vercors Programme Devoted to Safety Studies on Irradiated PWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tourasse, M.; Andre, B.; Ducros, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1996-12-31

    The first objective of the Heva-Vercors Program is to improve the data of fission product release and behaviour after an extensive fuel temperature increase and loss of integrity of the fuel elements that occur in case of severe PWR accident. The program is co-funded by the French Nuclear Protection and Safety Institute (IPSN) and Electricite de France (EDF). The experiments are conducted in a shielded cell of the French Grenoble Nuclear Centre. For these tests, industrial fuel from French PWR reactor plants is used. In order to rebuild the short lived fission product inventory, a reirradiation is performed in the experimental Siloe reactor, prior to the test. Eight tests have been conducted in the frame of the Heva Program up to 2370 K in the 1983-1988 period. The main outcomes of these studies were linked to the volatile fission product release. This program has been extended by the Vercors one with higher fuel temperature (2600 K) and improved instrumentation: gamma spectrometry, emission tomography, metallography, scanning electron microscopy, energy dispersive X-ray analysis, X-ray diffraction are some of the experimental techniques used for on line and post test characterization. The knowledge of the behavior of low volatile fission product has been significantly improved with the six Vercors tests. The results of the Vercors 4 test (38 GWd/t(U), 2570 K, reducing atmosphere) are presented here as an example. The key parameters are exhibited. The next step of these studies will use the Vercors HT loop that is planned to be operational at the beginning of 1996 to reach fuel melting temperature. The first aim of these future tests is to study the behaviour of non volatile and transuranic elements. An even more sophisticated instrumentation is implemented to reach the goal. The use of MOX fuel, the interaction between fission product aerosols and structural materials (Ag-In-Cd) and the fuel granulometry effect will be the next steps of the experimental program

  18. Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Serrano, M.

    2005-07-01

    Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

  19. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  20. Human comment dynamics in on-line social systems

    Science.gov (United States)

    Wu, Ye; Zhou, Changsong; Chen, Maoying; Xiao, Jinghua; Kurths, Jürgen

    2010-12-01

    Human comment is studied using data from ‘tianya’ which is one of the most popular on-line social systems in China. We found that the time interval between two consecutive comments on the same topic, called inter-event time, follows a power-law distribution. This result shows that there is no characteristic decay time on a topic. It allows for very long periods without comments that separate bursts of intensive comments. Furthermore, the frequency of a different ID commenting on a topic also follows a power-law distribution. It indicates that there are some “hubs” in the topic who lead the direction of the public opinion. Based on the personal comments habit, a model is introduced to explain these phenomena. The numerical simulations of the model fit well with the empirical results. Our findings are helpful for discovering regular patterns of human behavior in on-line society and the evolution of the public opinion on the virtual as well as real society.

  1. On-Line Core Thermal-Hydraulic Model Improvement

    International Nuclear Information System (INIS)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-01

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS

  2. Utilities enticing customers to come on-line

    International Nuclear Information System (INIS)

    Anon

    2000-01-01

    The first tentative steps by electric utilities to offer customer services on-line is reported. While most of the on-line communications to date has been merely to present information about products and services, at least a few utilities, -- Newfoundland Power being one of them -- are now offering customers the opportunity to check on their account status, to make inquiries, and on a voluntary basis employees of the utility can receive their bills on the web. BC Hydro is even more advanced; it has offered a similar service since 1997. The option to pay bills at the BC Hydro website is coming shortly. U. S. utility companies are reported to be far more advanced in the use of the Internet; according to a study by Deloitte Consulting, U.S. utilities are advancing to the next stage wherein Web intermediaries will be offering 'shop bots' that do comparison shopping on behalf of a customer, including bidding for power on a customer's behalf at energy auctions, reverse auctions, where sellers are bidding for customers' services, and buyers clubs where customers join together to take advantage of volume buying power

  3. On-line monitoring and inservice inspection in codes

    International Nuclear Information System (INIS)

    Bartonicek, J.; Zaiss, W.; Bath, H.R.

    1999-01-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

  4. Polar On-Line Acquisition Relay and Transmission System (POLARATS)

    Energy Technology Data Exchange (ETDEWEB)

    Yuracko, K.

    2004-07-15

    POLARATS (Polar On-Line Acquisition Relay And Transmission System) is being developed by YAHSGS LLC (YAHSGS) and Oak Ridge National Laboratory (ORNL) to provide remote, unattended monitoring of environmental parameters under harsh environmental conditions. In particular, instrumental design and engineering is oriented towards protection of human health in the Arctic, and with the additional goal of advancing Arctic education and research. POLARATS will obtain and transmit environmental data from hardened monitoring devices deployed in locations important to understanding atmospheric and aquatic pollutant migration as it is biomagnified in Arctic food chains. An Internet- and personal computer (PC)-based educational module will provide real time sensor data, on-line educational content, and will be integrated with workbooks and textbooks for use in middle and high school science programs. The educational elements of POLARATS include an Internet-based educational module that will instruct students in the use of the data and how those data fit into changing Arctic environments and food chains. POLARATS will: (1) Enable students, members of the community, and scientific researchers to monitor local environmental conditions in real time over the Internet; and (2) Provide additional educational benefits through integration with middle- and high-school science curricula. Information will be relayed from POLARATS devices to classrooms and libraries along with custom-designed POLARATS teaching materials that will be integrated into existing curricula to enhance the educational benefits realized from the information obtained.

  5. Aria Sardinia: the on line community joining tradition and innovatiom

    Directory of Open Access Journals (Sweden)

    Fabrizio Lao

    2005-12-01

    Full Text Available The "ARIA Sardinia" project (Network Actions for Italians Abroad has been especially designed to integrate and give value to the network of relationships between public administrations, local socio-economic stakeholders and Italian communities abroad, this goal to be pursued with the support of new technologies and learning approaches emerging within the context of on line interest communities. The general objective of the project is the development of competencies and knowledge, intended to combine specific technical skills with local "territorial knowledge", in a process where the strengthening and the dissemination of these forms of culture come from the prompt use of innovative tools. The main activities of the project are the actions intended to guide and assist entrepreneurs, associations, development projects' partners or promoters in the path of acquisition and dissemination of the competencies which are necessary to the involvement of Italians abroad into the internationalization process of Sardinian economy. ARIA Sardinia was funded by the Italian Foreign Affairs Ministry and the European Social Fund (FSE, within the framework of the National Operational Program for Technical Assistance and System Action (PON ATAS aimed at specific promotion initiatives and fostering of permanent links between Southern Italy economy and Italians living abroad. Keywords: on line community, networking, Italians abroad, Sardinian economy, Sardinia, culture.

  6. On-Line Core Thermal-Hydraulic Model Improvement

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan; Hwang, Dae Hyun; Seo, Kyung Won

    2007-02-15

    The objective of this project is to implement a fast-running 4-channel based code CETOP-D in an advanced reactor core protection calculator system(RCOPS). The part required for the on-line calculation of DNBR were extracted from the source of the CETOP-D code based on analysis of the CETOP-D code. The CETOP-D code was revised to maintain the input and output variables which are the same as in CPC DNBR module. Since the DNBR module performs a complex calculation, it is divided into sub-modules per major calculation step. The functional design requirements for the DNBR module is documented and the values of the database(DB) constants were decided. This project also developed a Fortran module(BEST) of the RCOPS Fortran Simulator and a computer code RCOPS-SDNBR to independently calculate DNBR. A test was also conducted to verify the functional design and DB of thermal-hydraulic model which is necessary to calculate the DNBR on-line in RCOPS. The DNBR margin is expected to increase by 2%-3% once the CETOP-D code is used to calculate the RCOPS DNBR. It should be noted that the final DNBR margin improvement could be determined in the future based on overall uncertainty analysis of the RCOPS.

  7. The Automated Threaded Fastening Based on On-line Identification

    Directory of Open Access Journals (Sweden)

    Nicolas Ivan Giannoccaro

    2008-11-01

    Full Text Available The principle of the thread fastenings have been known and used for decades with the purpose of joining one component to another. Threaded fastenings are popular because they permit easy disassembly for maintenance, repair, relocation and recycling. Screw insertions are typically carried out manually. It is a difficult problem to automat. As a result there is very little published research on automating threaded fastenings, and most research on automated assembly focus on the peg-in-hole assembly problem. This paper investigates the problem of automated monitoring of the screw insertion process. The monitoring problem deals with predicting integrity of a threaded insertion, based on the torque vs. insertion depth curve generated during the insertions. The authors have developed an analytical model to predict the torque signature signals during self-tapping screw insertions. However, the model requires parameters on the screw dimensions and plate material properties are difficult to measure. This paper presents a study on on-line identification during screw fastenings. An identification methodology for two unknown parameter estimation during a self-tapping screw insertion process is presented. It is shown that friction and screw properties required by the model can be reliably estimated on-line. Experimental results are presented to validate the identification procedure.

  8. Calibration through on-line monitoring of instruments channels

    International Nuclear Information System (INIS)

    James, R.W.

    1996-01-01

    Plant technical specifications require periodic calibration of instrument channels, and this has traditionally meant calibration at fixed time intervals for nearly all instruments. Experience has shown that unnecessarily frequent calibrations reduce channel availability and reliability, impact outage durations, and increase maintenance costs. An alternative approach to satisfying existing requirements for periodic calibration consists of on-line monitoring and quantitative comparison of instrument channels during operation to identify instrument degradation and failure. A Utility Working Group has been formed by EPRI to support the technical activities necessary to achieve generic NRC acceptance of on-line monitoring of redundant instrument channels as a basis for determining when to perform calibrations. A topical report proposing NRC acceptance of this approach was submitted in August 1995, and the Working Group is currently resolving NRC technical questions. This paper describes the proposed approach and the current status of the topical report with regard to NRC review. While these activities will not preclude utilities from continuing to use existing calibration approaches, successful acceptance of this performance-based approach will allow utilities to substantially reduce the number of calibrations which are performed. Concurrent benefits will include reduced I ampersand C impact on outage durations and improved sensitivity to instrument channel performance

  9. On-line optimal control improves gas processing

    International Nuclear Information System (INIS)

    Berkowitz, P.N.; Papadopoulos, M.N.

    1992-01-01

    This paper reports that the authors' companies jointly funded the first phase of a gas processing liquids optimization project that has the specific purposes to: Improve the return of processing natural gas liquids, Develop sets of control algorithms, Make available a low-cost solution suitable for small to medium-sized gas processing plants, Test and demonstrate the feasibility of line control. The ARCO Willard CO 2 gas recovery processing plant was chosen as the initial test site to demonstrate the application of multivariable on-line optimal control. One objective of this project is to support an R ampersand D effort to provide a standardized solution to the various types of gas processing plants in the U.S. Processes involved in these gas plants include cryogenic separations, demethanization, lean oil absorption, fractionation and gas treating. Next, the proposed solutions had to be simple yet comprehensive enough to allow an operator to maintain product specifications while operating over a wide range of gas input flow and composition. This had to be a supervisors system that remained on-line more than 95% of the time, and achieved reduced plant operating variability and improved variable cost control. It took more than a year to study various gas processes and to develop a control approach before a real application was finally exercised. An initial process for C 2 and CO 2 recoveries was chosen

  10. Amdel on-line analyser at Rooiberg Tin Limited

    International Nuclear Information System (INIS)

    Owen, T.V.

    1987-01-01

    An Amdel on line analysis system was installed on the 'A' mine tin flotation plant at Rooiberg in April 1984. The motivation for the installation was made on account of the large variations in the feed grade to the plant and the resulting need for rapid operational adjustments to control concentrate grades thereby maximising the financial returns. An 'on-line' analyser system presented itself as a suitable alternative to the existing control method of smaller laboratory x-ray fluorescence analysers. On the system as installed at Rooiberg, two probes were fitted in each analysis zone, viz a density probe using high energy gamma radiation from a Cesium 127 source and a specific element absorption probe using low energy gamma radiation from a Americium 241 source. The signals as received from the probes are fed to a line receiver unit in the control room where a micro computer is doing the processing and prints out the information as required. Several advantages of this type of installation were gained at Rooiberg Tin Limited

  11. On-line Social Interactions and Executive Functions

    Directory of Open Access Journals (Sweden)

    Oscar eYbarra

    2012-04-01

    Full Text Available A successful social interaction requires fast, on-line, and active construction of an ever-changing mental-model of another’s person beliefs, expectations, emotions, and desires. It also requires the ability to inhibit inappropriate behaviors, problem-solve, take-turns, and pursue goals in a distraction-rich environment. All these tasks rely on executive functions (EF—working memory, attention/cognitive control, and inhibition. Executive functioning has long been viewed as relatively static. However, starting with recent reports of successful cognitive interventions, this view is changing and now EFs are seen as much more open to both short and long term training, warm-up, and exhaustion effects. Some of the most intriguing evidence comes from research showing how social interaction enhances performance on standard EF tests. Interestingly, the latest research indicates these EF benefits are selectively conferred by certain on-line, dynamic social interactions, which require participants to engage with another person and actively construct the model of their mind. We review this literature and highlight its connection with evolutionary and cultural theories emphasizing links between intelligence and social life.

  12. Project development and commercialisation of on-line analysis systems

    International Nuclear Information System (INIS)

    Watt, J.S.

    2000-01-01

    A project team first in the Australian Atomic Energy Commission (AAEC) and since 1982 in CSIRO has developed many on-line analysis systems for the mineral and energy industries. The development of these projects, usually lasting 7-10 years, has followed a common pattern of laboratory R and D, field trials, commercialisation and technology transfer. This successful pattern is illustrated using examples of the development of systems for the on-line analysis of mineral slurries, for determination of the ash content of coal on conveyors, and for determination of the flow rates of oil, water and gas in pipelines. The first two systems, licensed to Australian companies, are used world-wide. They are now the market leaders for radioisotope gauges in their application field. The third, the multiphase flow meter, was licensed in 1997 to an international company. This meter has even greater potential than the other two systems for economic benefit from its use and for numbers of installations. (author)

  13. Increased Cortical Thickness in Professional On-Line Gamers

    Science.gov (United States)

    Hyun, Gi Jung; Shin, Yong Wook; Kim, Bung-Nyun; Cheong, Jae Hoon; Jin, Seong Nam

    2013-01-01

    Objective The bulk of recent studies have tested whether video games change the brain in terms of activity and cortical volume. However, such studies are limited by several factors including cross-sectional comparisons, co-morbidity, and short-term follow-up periods. In the present study, we hypothesized that cognitive flexibility and the volume of brain cortex would be correlated with the career length of on-line pro-gamers. Methods High-resolution magnetic resonance scans were acquired in twenty-three pro-gamers recruited from StarCraft pro-game teams. We measured cortical thickness in each individual using FreeSurfer and the cortical thickness was correlated with the career length and the performance of the pro-gamers. Results Career length was positively correlated with cortical thickness in three brain regions: right superior frontal gyrus, right superior parietal gyrus, and right precentral gyrus. Additionally, increased cortical thickness in the prefrontal cortex was correlated with winning rates of the pro-game league. Increased cortical thickness in the prefrontal and parietal cortices was also associated with higher performance of Wisconsin Card Sorting Test. Conclusion Our results suggest that in individuals without pathologic conditions, regular, long-term playing of on-line games is associated with volume changes in the prefrontal and parietal cortices, which are associated with cognitive flexibility. PMID:24474988

  14. Development of an artificial neural network model for on-line thermal margin estimation of a nuclear reactor core

    International Nuclear Information System (INIS)

    Kim, Hyun Koon

    1992-02-01

    One of the key safety parameters related to thermal margin in a Pressurized Water Reactor (PWR) core, is Departure from Nucleate Boiling Ratio (DNBR), which is to be assessed and continuously monitored during operation via either an analog or a digital monitoring system. The digital monitoring system, in general, allows more thermal margin than the analog system through the on-line computation of DNBR using the measured parameters as inputs to a simplified, fast running computer code. The purpose of this thesis is to develop an advanced method for on-line DNBR estimation by introducing an artifactual neural network model for best-estimation of DNBR at the given reactor operating conditions. the neural network model, consisting of three layers with five operating parameters in the input layer, provides real-time prediction accuracy of DNBR by training the network against the detailed simulation results for various operating conditions. The overall training procedure is developed to learn the characteristics of DNBR behaviour in the reactor core. First, a set of random combination of input variables is generated by Latin Hypercube Sampling technique performed on a wide range of input parameters. Second, the target values of DNBR to be referenced for training are calculated using a detailed simulation code, COBRA-IV. Third, the optimized training input data are selected. Then, training is performed using an Error Back Propagation algorithm. After completion of training, the network is tested on the examining data set in order to investigate the generalization capability of the network responses for the steady state operating condition as well as for the transient situations where DNB is of a primary concern. The test results show that the values of DNBR predicted by the neural network are maintained at a high level of accuracy for the steady state condition, and are in good agreements with the transient situation, although slightly conservative as compared to those

  15. On-line determination of moisture in coal and coke

    International Nuclear Information System (INIS)

    Cutmore, N.G.; Sowerby, B.D.

    1987-01-01

    The CSIRO Division of Mineral Engineering is developing various techniques for the on-line determination of moisture in coal and coke, and some instruments are now commercially available. These techniques permit accurate and rapid determination of moisture in materials directly on conveyor belts or in bins. The most promising techniques for direct on-belt measurement of moisture in coal are capacitance and microwave transmission. A non-contacting under-belt capacitance and gamma-ray backscatter technique has determined moisture in coal to better than 0.5 wt% in field tests. CSIRO is developing a fast neutron and gamma-ray transmission technique, which is proving very accurate in laboratory tests. This technique overcomes many of the limitations of thermal neutrons moisture gauges

  16. On-line monitoring system for utility boiler diagnostics

    International Nuclear Information System (INIS)

    Radovanovic, P.M.; Afgan, N.H.; Caralho, M.G.

    1997-01-01

    The paper deals with the new developed modular type Monitoring System for Utility Boiler Diagnostics. Each module is intended to assess the specific process and can be used as a stand alone application. Four modules are developed, namely: LTC - module for the on-line monitoring of parameters related to the life-time consumption of selected boiler components; TRD - module for the tube rupture detection by the position and working fluid Ieakage quantity; FAM - module for the boiler surfaces fouling (slagging) assessment and FLAP - module for visualization of the boiler furnace flame position. All four modules are tested on respective pilot plants built oil the 200 and 300 MWe utility boilers. Monitoring System is commercially available and can be realized in any combination of its modules depending on demands induced by the operational problems of specific boiler. Further development of Monitoring System is performed in accordance with the respective EU project on development of Boiler Expert System. (Author)

  17. On-line Certification for All: The PINVOX Algorithm

    Directory of Open Access Journals (Sweden)

    E Canessa

    2012-09-01

    Full Text Available A protoype algorithm: PINVOX (“Personal Identification Number by Voice" for on-line certification is introduced to guarantee that scholars have followed, i.e., listened and watched, a complete recorded lecture with the option of earning a certificate or diploma of completion after remotely attending courses. It is based on the injection of unique, randomly selected and pre-recorded integer numbers (or single letters or words within the audio trace of a video stream at places where silence is automatically detected. The certificate of completion or “virtual attendance” is generated on-the-fly after the successful identification of the embedded PINVOX code by a video viewer student.

  18. On-line control of the nonlinear dynamics for synchrotrons

    Science.gov (United States)

    Bengtsson, J.; Martin, I. P. S.; Rowland, J. H.; Bartolini, R.

    2015-07-01

    We propose a simple approach to the on-line control of the nonlinear dynamics in storage rings, based on compensation of the nonlinear resonance driving terms using beam losses as the main indicator of the strength of a resonance. The correction scheme is built on the analysis of the resonance driving terms in first perturbative order and on the possibility of using independent power supplies in the sextupole magnets, which is nowadays present in many synchrotron light sources. Such freedom allows the definition of "smart sextupole knobs" attacking each resonance separately. The compensation scheme has been tested at the Diamond light source and proved to be effective in opening up the betatron tune space, resonance free, available to the electron beam and to improve the beam lifetime.

  19. On-line control of the nonlinear dynamics for synchrotrons

    Directory of Open Access Journals (Sweden)

    J. Bengtsson

    2015-07-01

    Full Text Available We propose a simple approach to the on-line control of the nonlinear dynamics in storage rings, based on compensation of the nonlinear resonance driving terms using beam losses as the main indicator of the strength of a resonance. The correction scheme is built on the analysis of the resonance driving terms in first perturbative order and on the possibility of using independent power supplies in the sextupole magnets, which is nowadays present in many synchrotron light sources. Such freedom allows the definition of “smart sextupole knobs” attacking each resonance separately. The compensation scheme has been tested at the Diamond light source and proved to be effective in opening up the betatron tune space, resonance free, available to the electron beam and to improve the beam lifetime.

  20. On-line monitoring system for I-131 manufacturing labs

    International Nuclear Information System (INIS)

    Osovizky, A.; Malamud, Y.; Paran, Y.; Tal, N.; Turgeman, S.; Weinstein, M.

    1997-01-01

    An on-line monitoring and safety system has been installed in a lab for manufacturing 1-131 capsules for nuclear medicine use. Production of up to 100mCi batches is performed in shielded glove boxes. The safety system is based on a unique, 'Medi SMARTS' system (Medical Survey Mapping Automatic Radiation Tracing System), that collects continuously the radiation measurements for processing, display, and storage for future retrieval. Radiation is measured by GM tubes, data is transferred to a data processing unit, and then via a RS-485 communication line to a computer. In addition to the operational advantages and radiation levels storage, the system is being evaluated for the purpose of identifying risky stages in the process. (authors)