Operation of the main feedwater system turbopump following plant trip with total failure of the auxiliary feedwater system

International Nuclear Information System (INIS)

Lucas Alvaro, A.M. de; Rosa Martinez, B. de la; Alcaide, F.; Toledano Camara, C.

1993-01-01

The Auxiliary Feedwater System (AF) is a safeguard system which has been designed to supply feedwater to the steam generators, cool the primary system and remove decay heat from the reactor when the main feedwater pumps fail due to loss of power or any other reason. Thus, when plant trip occurs, the AF system pumps start up automatically, allowing removal of decay heat from the reactor. However, even though this system (2 motor-driven pumps and 1 turbopump) is highly reliable, injection of water to the steam generators must be ensured when it fails completely. To do this, if plant trip has not been caused by loss of off site power or failure of the Main Feedwater System (FW) turbopumps, one of these turbopumps can be used to achieve removal of decay heat. Since a large amount of steam is consumed by these turbopumps, an analysis has been performed to determine whether one of these pumps can be used and what actions are necessary to inject water into the steam generators. Results show that, for the case in question, a FW turbopump can be used to remove decay heat from the reactor. (author)

Reliability analysis of the auxiliary feedwater system; Analiza zanesljivosti sistema pomozne napajalne vode

Energy Technology Data Exchange (ETDEWEB)

Susnik, J; Dusic, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1984-07-01

The reliability of a NPP auxiliary feedwater system is evaluated using the fault tree analysis. The system is analyzed during the time interval 0 to 6 hours with the computer package program PREP/KITT which is described in more detail. (author)

Common-cause failure analysis of McGuire Unit 2 auxiliary feedwater system

International Nuclear Information System (INIS)

Rasmuson, D.M.; Shepherd, J.C.; Fowler, R.D.; Summitt, R.L.; Logan, B.W.

1982-01-01

A powerful method for qualitative common cause failure analysis (CCFA) of nuclear power plant systems was developed by EG and G Idaho at the Idaho National Engineering Laboratory. As a cooperative project to demonstrate and evaluate the usefulness of the method, the Duke Power Company agreed to allow a CCFA of the auxiliary feedwater system (AFWS) in their McGuire Nuclear Station Unit 2. The results of the CCFA are the subject of this discussion

Aging assessment of auxiliary feedwater pumps

International Nuclear Information System (INIS)

Greenstreet, W.L.

1987-01-01

ORNL is conducting aging assessments of auxiliary feedwater pumps to provide recommendations for monitoring and assessing the severity of time-dependent degradation as well as to recommend maintenance and replacement practices. Cornerstones of these activities are the identification of failure modes and causes and ranking of causes. Failure modes and causes of interest are those due to aging and service wear. Design details, functional requirements, and operating experience data were used to identify failure modes and causes and to rank the latter. Based on this input, potentially useful inspection, surveillance, and condition monitoring methods that are currently available for use or in the developmental stage were examined and recommendations made. The methods selected are listed and discussed in terms of use and information to be obtained. Relationships between inspection, surveillance, and monitoring and maintenance practices entered prominently into maintenance recommendations. These recommendations, therefore, embrace predictive as well as corrective and preventative maintenance practices. The recommendations are described, inspection details are discussed, and periodic inspection and maintenance interval guidelines are given. Surveillance testing at low-flow conditions is also discussed. It is shown that this type of testing can lead to accelerated aging

San Onofre/Zion auxiliary feedwater system seismic fault tree modeling

International Nuclear Information System (INIS)

Najafi, B.; Eide, S.

1982-02-01

As part of the study for the seismic evaluation of the San Onofre Unit 1 Auxiliary Feedwater System (AFWS), a fault tree model was developed capable of handling the effect of structural failure of the plant (in the event of an earthquake) on the availability of the AFWS. A compatible fault tree model was developed for the Zion Unit 1 AFWS in order to compare the results of the two systems. It was concluded that if a single failure of the San Onofre Unit 1 AFWS is to be prevented, some weight existing, locally operated locked open manual valves have to be used for isolation of a rupture in specific parts of the AFWS pipings

Probabilistic analysis of reactor safety - The auxiliary feedwater system of Angra I

International Nuclear Information System (INIS)

Oliveira, L.C.R. da L.C. de.

1981-09-01

The unavailability of the auxiliary feedwater system (AFWS) of Angra-1, was calculated. The fault tree analysis technique was used, considering two diferent types of contribution to system unavailability: The one due to hard-ware failure and the contribution due to test and maintenance which was separately analysed. The COMBO-and SAMPLE computer codes were used. The results have shown that the AFWS of Angra-1 contains enough redundancy to guarantee a safe operation under the conditions analysed, best values having been obtained for the unavailability of AFWS of Angra 1 with those codes than with the WASH-1400. (E.G.) [pt

Aging and service wear of auxiliary feedwater pumps for PWR nuclear power plants

International Nuclear Information System (INIS)

Greenstreet, W.L.

1989-01-01

This paper describes investigations on auxiliary feedwater pumps being done under the Nuclear Plant Aging Research (NPAR) Program. Objectives of these studies are: to identify and evaluate practical, cost-effective methods for detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear); recommend inspection and maintenance practices; establish acceptance criteria; and help facilitate use of the results. Emphasis is given to identifying and assessing methods for detecting failure in the incipient stage and to developing degradation trends to allow timely maintenance, repair or replacement actions. 3 refs

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

A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

International Nuclear Information System (INIS)

Svenson, Ola

1998-01-01

This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant

International Nuclear Information System (INIS)

Araujo, Jefferson Borges

1998-01-01

The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

Comparison of auxiliary feedwater and EDRS operation during natural circulation of MRX

International Nuclear Information System (INIS)

Kim, Jae Hak; Park, Goon Cherl

1997-01-01

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control rod drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation

Reliability analysis of 2 types of auxiliary feedwater system for PWR

International Nuclear Information System (INIS)

Ekariansyah, Andi Sofrany

2002-01-01

This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

International Nuclear Information System (INIS)

Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1992-10-01

In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant

Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

International Nuclear Information System (INIS)

Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E.

1991-09-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab

Steam generator auxiliary systems

International Nuclear Information System (INIS)

Heinz, A.

1982-01-01

The author deals with damage and defect types obtaining in auxiliary systems of power plants. These concern water/steam auxiliary systems (feed-water tank, injection-control valves, slide valves) and air/fluegas auxiliary systems (blowers, air preheaters, etc.). Operating errors and associated damage are not dealt with; by contrast, weak spots are pointed out which result from planning and design. Damage types and events are collected in statistics in order to facilitate damage evaluation for arriving at improved design solutions. (HAG) [de

Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Palo Verde was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Palo Verde plants

Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1994-05-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant

Auxiliary feedwater system risk-based inspection guide for the Maine Yankee Nuclear Power Plant

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Moffitt, N.E.; Bumgardner, J.D.

1992-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. The information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Maine Yankee was selected as one of a series of plants for study. ne product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Maine Yankee plant

Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

International Nuclear Information System (INIS)

Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Vehec, T.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant

Reliability analysis of the auxiliary feedwater system of Angra-1 including common cause failures using the multiple greek letter model

International Nuclear Information System (INIS)

Lapa, Celso Marcelo Franklin.

1996-05-01

The use of redundancy to increase the reliability of industrial systems make them subject to the occurrence of common cause events. The industrial experience and the results of safety analysis studies have indicated that common cause failures are the main contributors to the unreliability of plants that have redundant systems, specially in nuclear power plants. In this Thesis procedures are developed in order to include the impact of common cause failures in the calculation of the top event occurrence probability of the Auxiliary Feedwater System in a typical two-loop Nuclear Power Plant (PWR). For this purpose the Multiple Greek Letter Model is used. (author). 14 refs., 10 figs., 11 tabs

Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1993-12-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant

Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

International Nuclear Information System (INIS)

Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1991-07-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab

Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

International Nuclear Information System (INIS)

Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

1993-08-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant

Auxiliary feedwater system risk-based inspection guide for the J.M. Farley Nuclear Power Plant

International Nuclear Information System (INIS)

Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G.

1990-10-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment(PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab

A niching genetic algorithm applied to a nuclear power plant auxiliary feedwater system surveillance tests policy optimization

International Nuclear Information System (INIS)

Sacco, W.F.; Lapa, Celso M.F.; Pereira, C.M.N.A.; Oliveira, C.R.E. de

2006-01-01

This article extends previous efforts on genetic algorithms (GAs) applied to a nuclear power plant (NPP) auxiliary feedwater system (AFWS) surveillance tests policy optimization. We introduce the application of a niching genetic algorithm (NGA) to this problem and compare its performance to previous results. The NGA maintains a populational diversity during the search process, thus promoting a greater exploration of the search space. The optimization problem consists in maximizing the system's average availability for a given period of time, considering realistic features such as: (i) aging effects on standby components during the tests; (ii) revealing failures in the tests implies on corrective maintenance, increasing outage times; (iii) components have distinct test parameters (outage time, aging factors, etc.) and (iv) tests are not necessarily periodic. We find that the NGA performs better than the conventional GA and the island GA due to a greater exploration of the search space

Aging and low-flow degradation of auxilary feedwater pumps

International Nuclear Information System (INIS)

Adams, M.L.

1992-01-01

This paper documents the results of research done under the auspices of the Nuclear Regulatory Commission Nuclear Plant Aging Research Program. It examines the degradation imparted to safety related Auxiliary Feedwater System pumps at nuclear plants due to the low flow operation. The Auxiliary Feedwater (AFW) System is normally a stand-by system. As such it is operated most often in the test mode. Since few plants are equipped with full flow test loops, most testing is accomplished at minimum flow conditions in pump by-pass lines. It is the vibration and hydraulic forces generated at low flow conditions that have been shown to be the major causes of AFW pump aging and degradation. The wear can be manifested in a number of ways, such as impeller or diffuser breakage, thrust bearing and/or balance device failure due to excessive loading, cavitation damage on such stage impellers, increase seal leakage or failure, sear injection piping failure, shaft or coupling breakage, and rotating element seizure

Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

International Nuclear Information System (INIS)

Zhao Xiaoyu

1996-01-01

The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

Incorporation of passive components aging into PRAs

International Nuclear Information System (INIS)

Phillips, J.H.; Roesener, W.S.; Magleby, H.L.; Geidl, V.

1993-01-01

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants. (orig.)

Parallel island genetic algorithm applied to a nuclear power plant auxiliary feedwater system surveillance tests policy optimization

International Nuclear Information System (INIS)

Pereira, Claudio M.N.A.; Lapa, Celso M.F.

2003-01-01

In this work, we focus the application of an Island Genetic Algorithm (IGA), a coarse-grained parallel genetic algorithm (PGA) model, to a Nuclear Power Plant (NPP) Auxiliary Feedwater System (AFWS) surveillance tests policy optimization. Here, the main objective is to outline, by means of comparisons, the advantages of the IGA over the simple (non-parallel) genetic algorithm (GA), which has been successfully applied in the solution of such kind of problem. The goal of the optimization is to maximize the system's average availability for a given period of time, considering realistic features such as: i) aging effects on standby components during the tests; ii) revealing failures in the tests implies on corrective maintenance, increasing outage times; iii) components have distinct test parameters (outage time, aging factors, etc.) and iv) tests are not necessarily periodic. In our experiments, which were made in a cluster comprised by 8 1-GHz personal computers, we could clearly observe gains not only in the computational time, which reduced linearly with the number of computers, but in the optimization outcome

Loss-of-feedwater transients in PWRs

International Nuclear Information System (INIS)

Burns, R.D. III.

1980-01-01

Recent severe accident sequence analysis (SASA) work in LASL's Multifault Accident Analysis Section has focused on loss-of-feedwater (LOFW) transients at a 4-loop Westinghouse nuclear power reactor. In all transients studied, the initiator was loss of main feedwater and reactor coolant pump (RCP) trip, caused by temporary loss of off-site power. Subsequent automatic actions included reactor scram, closure of the main steam isolation valves, and initiation of auxiliary feedwater (AFW) flow. TRAC-PD2 calculations were designed to study the consequences of AFW delivery rates below the minimum specified in the emergency operating procedures (EOPs) for the reference 4-loop plant. Six types of LOFW scenarios have been studied, including (1) zero AFW availability (nominal case), (2) initially zero AFW but full recovery after 2 h, (3) zero AFW with steam generator (SG) atmospheric relief valve (ARV) malfunction, (4) zero AFW with high pressure charging flow initiated after 2 h, and (5) zero AFW with delay in reactor scram. Additional cases were considered to study the effects of uncertainties in pressurizer heater/spray operation, operator manual initiation of high pressure charging flow, reactor initial conditions, and RCP and power coastdown characteristics. Nominal case results, rationale for selections of other cases, and lessons learned are summarized

Probabilistic common cause failure modeling for auxiliary feedwater system after the introduction of flood barriers

International Nuclear Information System (INIS)

Zheng, Xiaoyu; Yamaguchi, Akira; Takata, Takashi

2013-01-01

Causal inference is capable of assessing common cause failure (CCF) events from the viewpoint of causes' risk significance. Authors proposed the alpha decomposition method for probabilistic CCF analysis, in which the classical alpha factor model and causal inference are integrated to conduct a quantitative assessment of causes' CCF risk significance. The alpha decomposition method includes a hybrid Bayesian network for revealing the relationship between component failures and potential causes, and a regression model in which CCF parameters (global alpha factors) are expressed by explanatory variables (causes' occurrence frequencies) and parameters (decomposed alpha factors). This article applies this method and associated databases needed to predict CCF parameters of auxiliary feedwater (AFW) system when defense barriers against internal flood are introduced. There is scarce operation data for functionally modified safety systems and the utilization of generic CCF databases is of unknown uncertainty. The alpha decomposition method has the potential of analyzing the CCF risk of modified AFW system reasonably based on generic CCF databases. Moreover, the sources of uncertainty in parameter estimation can be studied. An example is presented to demonstrate the process of applying Bayesian inference in the alpha decomposition process. The results show that the system-specific posterior distributions for CCF parameters can be predicted. (author)

The rise and fall of the passive auxiliary weorðan and strict Verb-Second in the history of English

NARCIS (Netherlands)

Postma, G.J.; Los, Bettelou; de Haan, Pieter

2015-01-01

The rise and fall of the passive auxiliary weorðan (WERDEN) in the history of English is investigated. We provide a new structural analysis of why and in what languages the passive diathesis can / cannot use the copula BE as auxiliary. We will do so in a comparative perspective within Germanic and

Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

International Nuclear Information System (INIS)

Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

International Nuclear Information System (INIS)

1985-07-01

On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program

Regulatory analysis for the resolution of Generic Issue 125.II.7 ''Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break''

International Nuclear Information System (INIS)

Basdekas, D.L.

1988-09-01

Generic Issue 125.II.7 addresses the concern related to the automatic isolation of auxiliary feedwater (AFW) to a steam generator with a broken steam or feedwater line. This regulatory analysis provides a quantitative assessment of the costs and benefits associated with the removal of the AFW automatic isolation and concludes that no new regulatory requirements are warranted. 21 refs., 7 tabs

Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

International Nuclear Information System (INIS)

Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes

2017-01-01

This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

Energy Technology Data Exchange (ETDEWEB)

Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes, E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

International Nuclear Information System (INIS)

Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

1992-01-01

U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

Feedwater control system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

Excessive swing of the feedwater in nuclear reactor power supply apparatus on the occurrence of a transient is suppressed by injecting an anticipatory compensating signal (δWsub(fw)) into the control for the feedwater. Typical overshoot occurs on removal of a large part of the load, the steam flow is reduced so that the conventional control system reduces the flow of feedwater. At the same time there is a reduction of feedwater level in the steam generator because of the collapse of the bubbles under increased steam pressure. By the time the control responds to the drop in level, the apparatus has begun to stabilize so that there is overshoot. The anticipatory signal is derived from the boiling power (BP) which is a function of the nuclear power (Qsub(N)) developed, the enthalpy of saturated water (hsub(s)) and the enthalpy of the feedwater injected into the steam generator (hsub(fw)). From the boiling power (BP) and the increment in steam pressure resulting from the transient an anticipatory increment of feedwater flow is derived. This increment is added to the other parameters controlling the feedwater. (author)

Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Grush, W.H.; Koizumi, Y.; Woerth, S.C.

1983-01-01

An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

Evaluation of a Design Concept for the Combined Air-water Passive Cooling PAFS+

International Nuclear Information System (INIS)

Bae, Sung Won; Kwon, Taesoon

2014-01-01

The APR+ system provides the Passive Auxiliary Feed-water System (PAFS) for the passive cooling capability. However, the current design requirement for working time for the PAFS is about 8 hours only. Thus, current working time of PAFS can not meet the required 72 hours cooling capability for the long term SBO situation. To meet the 72 hours cooling, the pool capacity should be almost 3∼4 times larger than that of current water cooling tank. In order to continue the PAFS operation for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility study on the combined passive air-water cooling system. Figure 1 and 2 show the conceptual difference of the PAFS and combined passive air-water cooling system, respectively. Simple performance evaluation of the passive air cooling heat exchanger has been conducted by the MARS calculation. For the postulated FLB scenario, 4800 heat exchanger tubes and 5 m/s air velocity are not sufficient to sustain the PCCT pool level for 72 hour cooling. Further works on the system design and performance enhancing plan are required to fulfill the 72 hours long term passive cooling

Loss-of-normal-feedwater sensitivity studies for AP600 behavior characterization

International Nuclear Information System (INIS)

Saiu, G.

1996-01-01

Activity concerning the development of a RELAP5/MOD3 model to simulate the Westinghouse Electric Corporation AP600 is summarized. The aim is to gain initial insight into the capability of RELAP5 to simulate the behavior of AP600 safety features. A-loss-of-normal-feedwater event is studied. Of the transients that must be investigated, this transient has been chosen to be one of the most relevant because the response of the AP600 to a loss-of-normal-feedwater event differs significantly from that of current pressurized water reactors in the extensive use of passive safety features peculiar to the AP600. Also, strong interactions among the AP600 safety systems, which should be further analyzed to permit full optimization of the system actuation logic and operation, are shown. Finally, a loss of normal feedwater without reactor scram, performed to investigate short-term plant behavior, shows that the pressure peak is affected by critical discharge flow coefficients applied to the pressurizer safety valves, while a relatively small reduction of the pressure peak is observed when both heat exchangers of the passive heat removal system are operating as opposed to the case in which only one is available. The data used for this study are derived from the Standard Safety Analysis Report configuration of the Westinghouse AP600 as of 1992

Feedwater control method and device therefor

International Nuclear Information System (INIS)

Nakahara, Mitsugu; Ichikawa, Yoshiaki; Ishii, Yoshikazu; Suzuki, Katsuyuki; Tanikawa, Naoshi; Mizuki, Fumio.

1997-01-01

The present invention provides a method of and a device for easily changing the constitution of feedwater systems without causing change in the water level of a reactor even when a plurality of feedwater systems have imbalance points. Namely, a feedwater control device comprises at least two feedwater systems capable of feeding water to tanks independently respectively and a controller capable of controlling water level in the tanks by controlling these feedwater systems. There is disposed a means for outputting gradually increasing driving signals to other feedwater systems, when the water level controller automatically controls one of the feedwater systems. There is also disposed a means for switching from automatic control for one of the feedwater systems to automatic control for the other feedwater system by a water level controller when the other feedwater system is in a stable operation region. As a result, entire feedwater flow rate is not temporarily changed and the water level in the tanks can be maintained constant. (N.H.)

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Auxiliary feedwater system risk-based inspection guide for the Diablo Canyon Unit 1 Nuclear Power Plant

International Nuclear Information System (INIS)

Gore, B.F.; Vo, T.V.; Harrison, D.G.

1990-08-01

This document presents a compilation of auxiliary feedwater (AFW) system failure information which has been screened for risk significance in terms of failure frequency and degradation of system performance. It is a risk-prioritized listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at Diablo Canyon. This information is presented to provide inspectors with increased resources for inspection planning at Diablo Canyon. The risk importance of various component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurized water reactors (PWRs). However, the component failure categories identified in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failures having a variety of root causes. In order to help inspectors to focus on specific aspects of component operation, maintenance and design which might cause these failures, an extensive review of component failure information was performed to identify and rank the root causes of these component failures. Both Diablo Canyon and industry-wide failure information was analyzed. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design problems, or component failures. This information permits an inspector to concentrate on components important to the prevention of core damage. Other components which perform essential functions, but which are not included because of high reliability or redundancy, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importances. 23 refs., 1 fig., 1 tab

Reactor feedwater pump control device

International Nuclear Information System (INIS)

Nishiyama, Hiroyuki.

1990-01-01

An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant; Um modelo de manutencao centrada em confiabilidade aplicada ao sistema de agua de alimentacaco auxiliar de uma usina nuclear

Energy Technology Data Exchange (ETDEWEB)

Araujo, Jefferson Borges

1998-01-15

The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network

International Nuclear Information System (INIS)

Lava, Deise Diana

2016-01-01

This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

Reactor feedwater device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

Purpose: To suppress soluble radioactive corrosion products in a feedwater device. Method: In a light water cooled nuclear reactor, an iron injection system is connected to feedwater pipeways and the iron concentration in the feedwater or reactor coolant is adjusted between twice and ten times of the nickel concentration. When the nickel/iron ratio in the reactor coolant or feedwater goes nearer to 1/2, iron ions are injected together with iron particles to the reactor coolant to suppress the leaching of stainless steels, decrease the nickel in water and increase the iron concentration. As a result, it is possible to suppress the intrusion of nickel as one of parent nuclide of radioactive nuclides. Further, since the iron particles intruded into the reactor constitute nuclei for capturing the radioactive nuclides to reduce the soluble radioactive corrosion products, the radioactive nuclides deposited uniformly to the inside of the pipeways in each of the coolant circuits can be reduced. (Kawakami, Y.)

Passive safety system of a super fast reactor

Energy Technology Data Exchange (ETDEWEB)

Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

2015-08-15

Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

The analysis of the functional role of man and machine in the control of a notional auxiliary feedwater system

International Nuclear Information System (INIS)

Cacciabue, P.C.; Codazzi, A.; Decortis, F.

1991-01-01

We will describe here the simulation of a moderately complex plant, i.e. the Auxiliary Feedwater System (AFWS) of a nuclear power plant, which has been developed for interacting with a cognitive model of operator in a simulation framework of man-machine system studies as well as with an external operator for verifying and validating the hypotheses of the theoretical model by experimental studies. In order to develop such simulation, which must be very flexible for satisfying the needs of interaction with an operator as well as with a cognitive model, a number of special conditions have been respected: the model of functional behaviour of the system has been extended to include the logic of control mechanisms, i.e. components, indicators and actuators; the control tasks for a number of sequences has been developed; the robustness of physical model has been tested in whole possible configuration of the plant; and finally, the interface of the simulation with the model for dynamic failures of components has also been granted. In this paper, these aspects of the deterministic model of the AFWS will be firstly presented in detail. Then, the interface of the plant simulation with an external user or with the cognitive model of the operator will be described focusing on the analysis of the control task. Finally, we will attempt to integrate our approach in an overall framework of taxonomy for studying human actions in complex work context

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Feedwater control system in nuclear power plants

International Nuclear Information System (INIS)

Masuyama, Hideo.

1981-01-01

Purpose: To enable switching operation for feedwater systems in a short time and with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of the fluctuations in the flow rate in the feedwater system during manual operation. Constitution: In a BWR type nuclear power plant having a plurality of feedwater systems to a nuclear reactor, a feedwater control system is constituted with a reactor water level controller, a M/A switcher for switching either of automatic flow rate demand signals or manual flow rate set signals from the reactor level controller to apply flow rate demand signals for each of the feedwater systems, a calculation device for calculating the flow rate set signals in the feedwater systems during manual operation and an adder for subtracting the flow rate set signals in the manual feedwater system calculated in the calculating device from the automatic flow rate demand signals for the feedwater systems during automatic operation. This enables rapid switching for the feedwater systems with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of fluctuations in the flow rate in the feedwater systems during manual operation and compensating the effects in upon manual and automatic switching by the M/A switcher. (Seki, T.)

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

Feedwater system in a nuclear power plant

International Nuclear Information System (INIS)

Shimizu, Tadayuki.

1975-01-01

Object: To improve the control property of a steam turbine for a feedwater pump and plant operation characteristics where water is supplied at a low rate. Structure: In a nuclear power plant where feedwater pumps of the reactor are driven by a steam turbine, the main feedwater duct on the discharge side of the feedwater pumps is provided with a cut-off valve and is connected parallel with a bypass duct having a pressure compensated flow control valve. With this arrangement, at the time when the rate of feedwater is high the cut-off valve is open so that water supplied from the feedwater pumps driven by the steam turbine is supplied through the main feedwater duct to the reactor while in case when the rate of feedwater is low the flow control valve is opened to let the water be supplied through the bypass duct. (Kamimura, M.)

Analysis of Total Loss of Feedwater for APR1400 using SPACE

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Min; Park, Seok Jeong; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction, Deajeon (Korea, Republic of)

2016-10-15

The Total Loss of FeedWater (TLOFW) event is an accident that main feedwater and auxiliary feedwater of secondary side are not supplied to steam generators. APR1400 uses the Safety Depressurization and Vent System (SDVS) for the F and B operation and SDVS is designed to perform the rapid depressurization function of Reactor Coolant System (RCS) through the remote manual operation when TLOFW is occurred. If RCS pressure falls below a Safety Injection Pump (SIP) working pressure, it can be possible to start the F and B operation which injects SIP flow to RCS and releases the RCS vapor and two-phase flow through Pilot Operated Safety Relief Valves (POSRVs) by opening the POSRVs, and then it can be possible to remove the decay heat. The design requirement of SDVS is that the core water level should be maintained at higher than 2 feet from the top of active core during the F and B operation. The TLOFW analysis was carried out to evaluate the capability of decay heat removal for APR1400 using newly developed SPACE code. The analysis results show that the F and B operation with 2 POSRVs and 2 SIPs and the F and B operation with 4 POSRVs and 4 SIPs meet the SDVS design requirement for the fuel cladding temperature. The comparison with RELAP5 shows good agreement and it validates the applicability of SPACE code for this type of accident analysis.

Reactor feedwater facility

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-30

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

49 CFR 230.57 - Injectors and feedwater pumps.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Injectors and feedwater pumps. 230.57 Section 230... Appurtenances Injectors, Feedwater Pumps, and Flue Plugs § 230.57 Injectors and feedwater pumps. (a) Water.... Injectors and feedwater pumps must be kept in good condition, free from scale, and must be tested at the...

Conceptual design of passive containment cooling system with air holdup tanks of improved APR+

International Nuclear Information System (INIS)

Jeon, Byong Guk; Cheon No, Hee

2014-01-01

In Korea, after the successful validation of passive auxiliary feedwater system (PAFS), a passive containment cooling system (PCCS) gets attention for future development. We suggested PCCS design based on APR+, an advanced PWR developed in Korea, and performed scoping analysis. On the extension of the simple scoping analysis, MARS simulation is performed to incorporate the behavior of water pool outside the containment as well as steam-air mixture inside the containment. Through the simulation we demonstrated the effectiveness of the air holdup tank (AHT). Also we investigated the effect of the models of heat transfer coefficients between steam-air mixture side and water side, and flow instability inside HX tubes. The presence of AHT enables us to reduce the number of required HX tubes more than half through an increase in the heat transfer coefficients due to the reduction of air fraction in the containment. Finally flow instability was observed and mitigated by putting orifice plates at the inlet of tubes, increasing height of return nozzle, and increasing a tube angle. (authors)

Interpretation of risk significance of passive component aging using probabilistic structural analysis

International Nuclear Information System (INIS)

Phillips, J.H.; Atwood, C.L.

1993-01-01

The probabilistic risk assessments (PRAs) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. Except as initiating events, the possible failure of passive components is given little consideration. The NRC is sponsoring a project at INEL to investigate the risk significance of passive components as they age. For this project, we developed a technique to calculate the failure probability of passive components over time, and demonstrated the technique by applying it to a weld in the auxiliary feedwater (AFW) system. A decreasing yearly rupture rate for this weld was calculated instead of the increasing rupture rate trend one might expect. We attribute this result to infant mortality; that is, most of those initial flaws that will eventually lead to rupture will do so early in life. This means that although each weld in a population may be wearing out, the population as a whole can exhibit a decreasing rupture rate. This observation has implications for passive components in commercial nuclear plants and other facilities where aging is a concern. For the population of passive components that exhibit a decreasing failure rate, risk increase is not a concern. The next step of the work is to identify the attributes that contribute to this decreasing rate, and to determine any attributes that would contribute to an increasing failure rate and thus to an increased risk

Development of methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWRs

International Nuclear Information System (INIS)

Shvarts, S.; Gerber, D.A.; House, K.; Hirschberg, P.

1994-01-01

The objective of this paper is to describe a methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWR plants. This methodology is based in part on plant test data obtained from a recent Diablo Canyon Power Plant (DCPP) Unit 1 heatup. Temperature sensors installed near the nozzle-to-pipe weld were monitored during the heatup, along with operational parameters such as auxiliary feedwater (AFW) flow rate and steam generator temperature. A thermal stratification load definition was developed from this data. Steady state characteristics of this data were used in a finite element analysis to develop relationship between AFW flow and stratification interface level. Fluctuating characteristics of this data were used to determine transient parameters through the application of a Green's Function approach. The thermal stratification load definition from the test data was used in a three-dimensional thermal stress analysis to determine stress cycling and consequent fatigue damage or crack growth during AFW flow fluctuations. The implementation of the developed methodology in the DCPP and Sequoyah Nuclear Plant (SNP) fatigue monitoring systems is described

A Smart Soft Sensor Predicting Feedwater Flow Rate

International Nuclear Information System (INIS)

Yang, Heon Young; Na, Man Gyun

2009-01-01

Since we evaluate thermal nuclear reactor power with secondary system calorimetric calculations based on feedwater flow rate measurements, we need to measure the feedwater flow rate accurately. The Venturi flow meters that are being used to measure the feedwater flow rate in most pressurized water reactors (PWRs) measure the flow rate by developing a differential pressure across a physical flow restriction. The differential pressure is then multiplied by a calibration factor that depends on various flow conditions in order to calculate the feedwater flow rate. The calibration factor is determined by the feedwater temperature and pressure. However, Venturi meters cause a buildup of corrosion products near the orifice of the meter. This fouling increases the measured pressure drop across the meter, thereby causing an overestimation of the feedwater flow rate

Feedwater device for nuclear power plant

International Nuclear Information System (INIS)

Ikekita, Iwao.

1980-01-01

Purpose: To conduct water feeding without using high pressure steam of the reactor and with no radiation exposure by the provision of each feedwater pump driven by each motor controlled from variable frequency thyristor-inverter to a feedwater pipe connecting a condensate pump and the reactor. Constitution: High pressure steams resulted from heat exchange in the reactor core are transferred by way of a main steam check valve in a main steam pipe to a high pressure turbine, drive the high pressure turbine, flow out of the turbine and then drive a low pressure turbine by way of a moisture separator. The steams thus used for the turbine driving are condensed in a condensator and then sent under pressure by way of each condensating pump to a feedwater pipe. Since each of the feedwater pumps provided in the route of the feedwater pipe is driven by each of the motors under the control of the variable frequency thyristor-inverter in starting, shut down and normal operation, water is fed to the reactor. (Horiuchi, T.)

Feedwater recycling system in BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

Feedwater processing method in a boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Izumitani, M; Tanno, K

1976-09-06

The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

Preliminary design of RDE feedwater pump impeller

International Nuclear Information System (INIS)

Sri Sudadiyo

2018-01-01

Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

International Nuclear Information System (INIS)

Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

2015-01-01

Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

ESBWR power maneuvering via feedwater temperature control

International Nuclear Information System (INIS)

Saha, P.; Marquino, W.; Tucker, L. J.

2008-01-01

The ESBWR is a Generation III+ Boiling Water Reactor (BWR) driven by natural circulation. For a given geometry/hardware, system pressure, downcomer water level and feedwater temperature, the core flow rate in the ESBWR is only a function of reactor power, controlled through the control blade movement. In order to provide operational flexibility, another method of core-wide or global power maneuvering via feedwater temperature control has been developed. This is independent of power maneuvering via control blade movement, and it lowers the linear heat generation rate (LHGR) changes near the tip of control blades, which improves fuel reliability. All required stability, anticipated operational occurrences (AOOs), infrequent events, special events including anticipated transients without scram (ATWS), and loss-of-coolant accident (LOCA) analyses have been performed for the 4500 MWt ESBWR. Based on the results of these analyses at 'high', nominal and 'low' feedwater temperatures, a safe Power - Feedwater Temperature operating domain has been developed. This paper summarizes the results of these analyses and presents the ESBWR Power - Feedwater Temperature operating domain or map. (authors)

NucleDyne's passive containment system

International Nuclear Information System (INIS)

Falls, O.B. Jr.; Kleimola, F.W.

1987-01-01

A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

Feedwater heater

International Nuclear Information System (INIS)

Murata, Shigeto; Minato, Akihiko; Yokomizo, Osamu; Masuhara, Yasuhiro.

1991-01-01

The present invention concerns a feedwater heater for a BWR type reactor. A cylinder is fit into the lower portion of a drain inlet pipe, to which drain water inflows from a turbine, and a disk is disposed to the lower end of the cylinder vertically to the axis of the cylinder, to constitute a drain water dispersing mechanism. Drain water inflown from the drain inlet pipe is fallen in the cylinder and collides against the disk. The collided drain water is splashed horizontally by its kinetic energy to reach the heat transfer pipe and conducts heat exchange. In this case, the drain water is converted into fine droplets by the collision against the disk and scattered in a wide range in the heater. As a result, sensible heat in the drain water can be transferred to feedwater effectively. Then, even the heat energy of the drain water can be utilized effectively for heat exchange, to improve the heat exchange efficiency. (I.N.)

Lead corrosion and transport in simulated secondary feedwater

International Nuclear Information System (INIS)

McGarvey, G.B.; Ross, K.J.; McDougall, T.E.; Turner, C.W.

1998-01-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

Lead corrosion and transport in simulated secondary feedwater

Energy Technology Data Exchange (ETDEWEB)

McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

1998-07-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

Minimum throttling feedwater control in VVER-1000 and PWR NPPs

International Nuclear Information System (INIS)

Symkin, B.E.; Thaulez, F.

2004-01-01

This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

Lead corrosion and transport in simulated secondary feedwater

Energy Technology Data Exchange (ETDEWEB)

McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W

1999-07-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is caused by low levels of lead although few, if any, failures have been experimentally linked to lead when it is present in sub-parts per billion in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the 3 principal corrosion products in the secondary feedwater: magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values that are lower than the pH of the feedwater (9 to 10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces after different treatment conditions will be used to propose a model for the transport and probable fate of lead in the secondary feedwater system. (author)

IE Information Notice No. 85-75: Improperly installed instrumentation, inadequate quality control and inadequate postmodification testing

International Nuclear Information System (INIS)

Jordan, E.L.

1992-01-01

On June 10, 1985, the licensee informed the NRC Resident Inspector that for approximately 5 days LaSalle Unit 2 had been without the capability of automatic actuation of emergency core cooling (ECCS) and that for approximately 3 days during this period the plant had been without secondary containment integrity. The major cause of this condition was improper installation (the variable and reference legs were reversed) of the two reactor vessel level actuation switches which control Division 1 automatic depressurization system (ADS), low pressure core spray (LPCS), and reactor core isolation cooling (RCIC). On July 20, 1985, the Trojan Nuclear Power Plant tripped from 100% power because of a turbine trip that was caused by the loss of the unit auxiliary transformer. All systems functioned normally except that low suction pressure caused one auxiliary feedwater pump to trip and then the other auxiliary feedwater pump to trip after restart of the first auxiliary feedwater pump. The cause of the trips of the auxiliary feedwater pumps can be traced back to improper postmodification adjustment and inadequate postmodification testing following retrofit of environmentally qualified controllers for the auxiliary feedwater system. The auxiliary feedwater pump trips on low suction pressure were caused by excessive combined flow from the two auxiliary feedwater pumps that draw from a single header from the condensate storage tank. The flow control valves were open farther than required after new environmentally qualified controllers had been installed during a recent refueling outage

Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

International Nuclear Information System (INIS)

Nassersharif, B.

1986-01-01

A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

Feedwater temperature control methods and systems

Science.gov (United States)

Moen, Stephan Craig; Noonan, Jack Patrick; Saha, Pradip

2014-04-22

A system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is disclosed. The system, in accordance with an example embodiment of the present invention, may include a controller configured to control a power output level of the NCBWR by controlling a heating subsystem to adjust a temperature of feedwater flowing into an annulus of the NCBWR. The heating subsystem may include a steam diversion line configured to receive steam generated by a core of the NCBWR and a steam bypass valve configured to receive commands from the controller to control a flow of the steam in the steam diversion line, wherein the steam received by the steam diversion line has not passed through a turbine. Additional embodiments of the invention may include a feedwater bypass valve for controlling an amount of flow of the feedwater through a heater bypass line to the annulus.

Aiding operator performance at low power feedwater control

International Nuclear Information System (INIS)

Woods, D.D.

1986-01-01

Control of the feedwater system during low power operations (approximately 2% to 30% power) is a difficult task where poor performance (excessive trips) has a high cost to utilities. This paper describes several efforts in the human factors aspects of this task that are underway to improve feedwater control. A variety of knowledge acquisition techniques have been used to understand the details of what makes feedwater control at low power difficult and what knowledge and skill distinguishes expert operators at this task from less experienced ones. The results indicate that there are multiple factors that contribute to task difficulty

Simulation of the KAERI PASCAL Test with MARS-KS and TRACE Codes

Energy Technology Data Exchange (ETDEWEB)

Lee, Kyung Won; Cheong, Aeju; Shin, Andong; Cho, Min Ki [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

In order to validate the operational performance of the PAFS, KAERI has performed the experimental investigation using the PASCAL (PAFS Condensing heat removal Assessment Loop) facility. In this study, we simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. We simulated the KAERI PASCAL SS-540-P1 test with MARS-KS V1.4 and TRACE V5.0 p4 codes to assess the code predictability for the condensation heat transfer inside the passive auxiliary feedwater system. The calculated results of heat flux, inner wall surface temperature of the condensing tube, fluid temperature, and steam mass flow rate are compared with the experimental data. The result shows that the MARS-KS generally under-predict the heat fluxes. The TRACE over-predicts the heat flux at tube inlet region and under-predicts it at tube outlet region. The TRACE prediction shows larger amount of steam condensation by about 3% than the MARS-KS prediction.

Condensate and feedwater systems, pumps, and water chemistry. Volume seven

International Nuclear Information System (INIS)

Anon.

1986-01-01

Subject matter includes condensate and feedwater systems (general features of condensate and feedwater systems, condenser hotwell level control, condensate flow, feedwater flow), pumps (principles of fluid flow, types of pumps, centrifugal pumps, positive displacement pumps, jet pumps, pump operating characteristics) and water chemistry (water chemistry fundamentals, corrosion, scaling, radiochemistry, water chemistry control processes, water pretreatment, PWR water chemistry, BWR water chemistry, condenser circulating water chemistry

Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

International Nuclear Information System (INIS)

Boucher, T.J.; Owca, W.A.

1985-05-01

This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

Passive system with steam-water injector for emergency supply of NPP steam generators

International Nuclear Information System (INIS)

Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

2009-01-01

The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

Evaluation of the PRHRS Performance Degradation due to Non-Condensable Gas for the Small and Medium Reactor using MARS-KS code

International Nuclear Information System (INIS)

Kim, Sook Kwan; Sim, Suk Ku; Park, Ju Yeop; Seol, Kwang Won; Ryu, Yong Ho

2011-01-01

The effect of non-condensable gas on the performance of PRHRS (Passive Residual Heat Removal System) of the Small and Medium Reactor(SMR) was evaluated during a loss of flow event. Since the TMI accident in 1979, the passive systems have been considered in the advanced reactors as a feature of design improvement because the passive system simplifies the system and thus increases the reliability of the system. The Westinghouse received the design certification from the USNRC for the AP600 and AP1000 passive type pressurized water reactors. The APR+ under development by KEPCO considers the use of PAFS (Passive Auxiliary Feedwater System). And the PRHRS is adopted as a passive secondary heat removal system for the SMART (System-integrated Modular Advanced ReacTor)

Secondary coolant circuit operation tests: steam generator feedwater supply

International Nuclear Information System (INIS)

Beroux, M.

1985-01-01

No one important accident occurred during the start-up tests of the 1300MWe P4 series, concerning the feedwater system of steam generators (SG). This communication comments on some incidents, that the tests allowed to detect very soon and which had no consequences on the operation of units: 1) Water hammer in feedwater tubes, and incidents met in the emergency steam generator water supply circuit. The technological differences between SG 900 and 1300 are pointed out, and the measures taken to prevent this problem are presented. 2) Incidents met on the emergency feedwater supply circuit of steam generators; mechanical or functional modifications involved by these incidents [fr

Dependence of steam generator vibrations on feedwater pressure

International Nuclear Information System (INIS)

Sadilek, J.

1989-01-01

Vibration sensors are attached to the bottom of the steam generator jacket between the input and output primary circuit collectors. The effective vibration value is recorded daily. Several times higher vibrations were observed at irregular intervals; their causes were sought, and the relation between the steam generator vibrations measured at the bottom of its vessel and the feedwater pressure was established. The source of the vibrations was found to be in the feedwater tract of the steam generator. The feedwater tract is described and its hydraulic characteristics are given. Vibrations were measured on the S02 valve. It is concluded that vibrations can be eliminated by reducing the water pressure before the control valves and by replacing the control valves with ones with more suitable control characteristics. (E.J.). 3 figs., 1 tab., 3 refs

Smart Soft-Sensing for the Feedwater Flowrate at PWRs Using a GMDH Algorithm

Science.gov (United States)

Lim, Dong Hyuk; Lee, Sung Han; Na, Man Gyun

2010-02-01

The thermal reactor power in pressurized water reactors (PWRs) is typically assessed using secondary system calorimetric calculations based on accurate measurements of the feedwater flowrate. Therefore, precise measurements of the feedwater flowrate are essential. In most PWRs, Venturi meters are used to measure the feedwater flowrate. However, the fouling phenomena of the Venturi meter deteriorate the accuracy of the existing hardware sensors. Consequently, it is essential to resolve the inaccurate measurements of the feedwater flowrate. In this study, in order to estimate the feedwater flowrate online with high precision, a smart soft sensing model for monitoring the feedwater flowrate was developed using a group method of data handling (GMDH) algorithm combined with a sequential probability ratio test (SPRT). The uncertainty of the GMDH model was also analyzed. The proposed sensing and monitoring algorithm was verified using the acquired real plant data from Yonggwang Nuclear Power Plant Unit 3.

Application of neural networks to validation of feedwater flow rate in a nuclear power plant

International Nuclear Information System (INIS)

Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

1993-01-01

Feedwater flow rate measurement in nuclear power plants requires periodic calibration. This is due to the fact that the venturi surface condition of the feedwater flow rate sensor changes because of a chemical reaction between the surface coating material and the feedwater. Fouling of the venturi surface, due to this chemical reaction and the deposits of foreign materials, has been observed shortly after a clean venturi is put in operation. A fouled venturi causes an incorrect measurement of feedwater flow rate, which in turn results in an inaccurate calculation of the generated power. This paper presents two methods for verifying incipient and continuing fouling of the venturi of the feedwater flow rate sensors. Both methods are based on the use of a set of dissimilar process variables dynamically related to the feedwater flow rate variable. The first method uses a neural network to generate estimates of the feedwater flow rate readings. Agreement, within a given tolerance, of the feedwater flow rate instrument reading, and the corresponding neural network output establishes that the feedwater flow rate instrument is operating properly. The second method is similar to the first method except that the neural network predicts the core power which is calculated from measurements on the primary loop, rather than the feedwater flow rates. This core power is referred to the primary core power in this paper. A comparison of the power calculated from the feedwater flow measurements in the secondary loop, with the calculated and neural network predicted primary core power provides information from which it can be determined whether fouling is beginning to occur. The two methods were tested using data from the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant

Heat exchanger inventory cost optimization for power cycles with one feedwater heater

International Nuclear Information System (INIS)

Qureshi, Bilal Ahmed; Antar, Mohamed A.; Zubair, Syed M.

2014-01-01

Highlights: • Cost optimization of heat exchanger inventory in power cycles is investigated. • Analysis for an endoreversible power cycle with an open feedwater heater is shown. • Different constraints on the power cycle are investigated. • The constant heat addition scenario resulted in the lowest value of the cost function. - Abstract: Cost optimization of heat exchanger inventory in power cycles with one open feedwater heater is undertaken. In this regard, thermoeconomic analysis for an endoreversible power cycle with an open feedwater heater is shown. The scenarios of constant heat rejection and addition rates, power as well as rate of heat transfer in the open feedwater heater are studied. All cost functions displayed minima with respect to the high-side absolute temperature ratio (θ 1 ). In this case, the effect of the Carnot temperature ratio (Φ 1 ), absolute temperature ratio (ξ) and the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ) are qualitatively the same. Furthermore, the constant heat addition scenario resulted in the lowest value of the cost function. For variation of all cost functions, the smaller the value of the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ), lower the cost at the minima. As feedwater heater to hot end unit cost ratio decreases, the minimum total conductance required increases

An effect of downcomer feedwater fraction on steam generator performance with an axial flow economizer

International Nuclear Information System (INIS)

Jung, Byung Ryul; Park, Hu Shin; Chung, Duk Muk; Baik, Se Jin

2000-01-01

The effects of feedwater flow fraction introduced into the downcomer region have been evaluated in terms of steam generator performance based on the same steam generator thermal output for the Korea Standard Nuclear Power Plant (KSNP) steam generator. The KSNP steam generator design has an integral axial flow economizer which is designed such that most of the feedwater is introduced through the economizer region and only a portion of feedwater through the downcomer region. The feedwater flow introduced into the downcomer region is not normally controlled during the power operation. However, the actual feedwater fraction into the downcomer region may differ from the design flow depending on the as-built system and component characteristics. Investigated in this paper were the downcomer feedwater flow effects on the steam pressure, circulation ratio, internal void fraction and velocity distribution in the tube bundle region at the steady state operation using SAFE and ATHOS3 codes. The results show that the steam pressure increases and the resultant total feedwater flow increases with reducing the downcomer feedwater flow fraction for the same steam generator thermal output. The slight off-design condition of downcomer feedwater flow fraction renders no significant effect on the steam generator performance such as circulation ratios, steam qualities, void fractions and internal velocity distributions. The evaluation shows that the slight off-design downcomer feedwater flow fraction deviation up to ± 5% is acceptable for the steam generator performance

Feedwater heater performance evaluation using the heat exchanger workstation

International Nuclear Information System (INIS)

Ranganathan, K.M.; Singh, G.P.; Tsou, J.L.

1995-01-01

A Heat Exchanger Workstation (HEW) has been developed to monitor the condition of heat exchanging equipment power plants. HEW enables engineers to analyze thermal performance and failure events for power plant feedwater heaters. The software provides tools for heat balance calculation and performance analysis. It also contains an expert system that enables performance enhancement. The Operation and Maintenance (O ampersand M) reference module on CD-ROM for HEW will be available by the end of 1995. Future developments of HEW would result in Condenser Expert System (CONES) and Balance of Plant Expert System (BOPES). HEW consists of five tightly integrated applications: A Database system for heat exchanger data storage, a Diagrammer system for creating plant heat exchanger schematics and data display, a Performance Analyst system for analyzing and predicting heat exchanger performance, a Performance Advisor expert system for expertise on improving heat exchanger performance and a Water Calculator system for computing properties of steam and water. In this paper an analysis of a feedwater heater which has been off-line is used to demonstrate how HEW can analyze the performance of the feedwater heater train and provide an economic justification for either replacing or repairing the feedwater heater

On-line validation of feedwater flow rate in nuclear power plants using neural networks

International Nuclear Information System (INIS)

Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

1994-01-01

On-line calibration of feedwater flow rate measurement in nuclear power plants provides a continuous realistic value of feedwater flow rate. It also reduces the manpower required for periodic calibration needed due to the fouling and defouling of the venturi meter surface condition. This paper presents a method for on-line validation of feedwater flow rate in nuclear power plants. The method is an improvement of the previously developed method which is based on the use of a set of process variables dynamically related to the feedwater flow rate. The online measurements of this set of variables are used as inputs to a neural network to obtain an estimate of the feedwater flow rate reading. The difference between the on-line feedwater flow rate reading, and the neural network estimate establishes whether there is a need to apply a correction factor to the feedwater flow rate measurement for calculation of the actual reactor power. The method was applied to the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant. The venturi meters used for flow measurements are susceptible to frequent fouling that degrades their measurement accuracy. The fouling effects can cause an inaccuracy of up to 3% relative error in feedwater flow rate reading. A neural network, whose inputs were the readings of a set of reference instruments, was designed to predict both feedwater flow rates simultaneously. A multi-layer feedforward neural network employing the backpropagation algorithm was used. A number of neural network training tests were performed to obtain an optimum filtering technique of the input/output data of the neural networks. The result of the selection of the filtering technique was confirmed by numerous Fast Fourier Transform (FFT) tests. Training and testing were done on data from TMI-1 nuclear power plant. The results show that the neural network can predict the correct flow rates with an absolute relative error of less than 2%

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

2009-01-01

TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-05-15

MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

A technique of including the effect of aging of passive components in probabilistic risk assessments

International Nuclear Information System (INIS)

Phillips, J.H.; Weidenhamer, G.H.

1992-01-01

The probabilistic risk assessments (PRAS) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing methods for selecting risk-significant passive components and including them in PRAS. These methods provide effective ways to prioritize passive components for inspection, and where inspection reveals aging damage, mitigation or repair can be employed to reduce the likelihood of component failure. We demonstrated a method by selecting a weld in the auxiliary feedwater (AFW) system, basing our selection on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then modified and used the Piping Reliability Analysis Including Seismic Events (PRAISE) computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The PRAISE code was modified to include the effects of changing design material properties with age and changing stress cycles. The calculation included the effects of mechanical loads and thermal transients typical of the service loads for this piping design and the effects of thermal cycling caused by a leaking check valve. However, this particular calculation showed little change in low component failure probability and plant risk for 48 years of service. However, sensitivity studies showed that if the probability of component failure is high, the effect on plant risk is significant. The success of this demonstration shows that this method could be applied to nuclear power plants. The demonstration showed the method is too involved (PRAISE takes a long time to perform the calculation and the input information is extensive) for handling a large number of passive components and therefore simpler methods are needed

Considerations for surviving the loss of a main feedwater pump at full power

International Nuclear Information System (INIS)

Gaydos, K.A.; Calvo, R.; Conroy, P.W.; Klein, C.M.; Mellers, J.E.

1990-01-01

Today's economics dictate that nuclear power operational costs be contained by addressing frequently-occurring trips that might be minimized or avoided via specific upgrades. Much recent attention has focused on the significant percentage of plant trips related to feedwater flow regulation; however, another frequent feedwater-related trip stems from the loss of a single main feedwater pump while operating at high power levels, causing a plant trip on low steam generator water-level. This paper summarizes the results of several plant-specific studies that evaluate a unit's capabilities to consistently survive the loss of a main feedwater pump from full power, and outlines a methodology for analyzing this capability

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

Manual for investigation and correction of feedwater heater failures

International Nuclear Information System (INIS)

Bell, R.J.; Diaz-Tous, I.A.; Bartz, J.A.

1993-01-01

The Electric Power Research Institute (EPRI) has sponsored the development of a recently published manual which is designed to assist utility personnel in identifying and correcting closed feedwater heater problems. The main portion of the manual describes common failure modes, probable means of identifying root causes and appropriate corrective actions. These include materials selection, fabrication practices, design, normal/abnormal operation and maintenance. The manual appendices include various data, intended to aid those involved in monitoring and condition assessment of feedwater heaters. This paper contains a detailed overview of the manual content and suggested means for its efficient use by utility engineers and operations and maintenance personnel who are charged with the responsibilities of performing investigations to identify the root cause(s) of closed feedwater problems/failures and to provide appropriate corrective actions. 4 refs., 3 figs., 2 tabs

Pressurized water-reactor feedwater piping response to water hammer

International Nuclear Information System (INIS)

Arthur, D.

1978-03-01

The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

Assessment of Flow Instability in Passive Auxiliary Feedwater System (PAFS) Using RELAP5

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cheon, Jong; Kim, Han-Gon [KHNP, Daejeon (Korea, Republic of)

2015-10-15

In this study, the occurrence possibility of both instabilities in PAFS is assessed with the best-estimate thermal hydraulic code, RELAP5. From the RELAP5 code analysis, the Ledinegg instability might not occur in PAFS. The DWO might occur in PAFS but the effect of the oscillation on the heat removal capacity of PAFS was not large. Therefore, it is concluded that PAFS is safe in terms of flow instabilities. Since PAFS is two-phase flow system, flow instabilities may occur. Flow instabilities may cause the severe deterioration of heat removal capability of PAFS due to the reduction of the condensate flow. For the reliable operation of PAFS, it is required to assess the flow instabilities in PAFS. The Ledinegg-type instability and the Density Wave Oscillation (DWO) are the representative static flow instability and the dynamic flow instability, respectively.

Water hammer calculation and analysis in main feedwater system of PWR nuclear power plants

International Nuclear Information System (INIS)

Wang Xin; Han Weishi

2010-01-01

The main feedwater system of a nuclear power plant is an important part in ensuring the cooling of the steam generator. Moreover, it is the main pipe section where water hammers frequently occur. Studying the regular patterns of water hammers to the main feedwater system is significant to the stable operation of the system. The paper focuses on the study of water hammers through Flowmaster's transient calculating function to establish a mathematical model with boundary conditions such as a feedwater pump, control valves, etc.; calculation of the water hammers pressure when feedwater pumps and control valves shut down; exporting the instantaneous change in solution of pressure. Combined with engineering practical examples, the conclusions verify the viability of calculating the water hammers pressure through Flowmaster's transient function, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively, changing the intervals of closing signals to feedwater pumps and control valves to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (authors)

Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

International Nuclear Information System (INIS)

Fuller, R.; Harrell, J.

1996-01-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

Energy Technology Data Exchange (ETDEWEB)

Fuller, R.; Harrell, J.

1996-12-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

International Nuclear Information System (INIS)

Anon.

1982-01-01

A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

A ZVT–ZCT PWM synchronous buck converter with a simple passive auxiliary circuit for reduction of losses and efficiency enhancement

Directory of Open Access Journals (Sweden)

S. Shiva Kumar

2015-06-01

Full Text Available In this paper, a Zero-Voltage-Transition (ZVT–Zero-Current Transition (ZCT Pulse-width Modulated (PWM synchronous buck converter (SBC, with a simple passive auxiliary circuit is proposed, which reduces the stresses and improves the efficiency by pacifying the conduction losses compared to a traditional PWM converter, typically suitable for photovoltaic applications. The important design feature of ZVT–ZCT PWM SBC converters is placement of resonant components that mollifies the conduction losses. Due to the ZVT–ZCT, the resonant components with low values are used, thereby resulting in the increase of switching frequency. The ZVT–ZCT operation of the proposed converter is presented through theoretical analysis. The characteristics of the proposed converter are verified with the simulation in the PSIM co-simulated with MATLAB/SIMULINK environment and implemented experimentally.

Loss of feedwater heater analysis for the South Texas Project

International Nuclear Information System (INIS)

Joyce, K.C.; Johnson, M.R.; Albury, C.R.

1987-01-01

The results of the steady state and transient analyses of the low pressure feedwater heater train for the South Texas Nuclear Project are presented. The South Texas Project consists of two 1250 MW Westinghouse PWR units. This analysis was performed using the Modular Modeling System (MMS) simulation code. The model presented will be incorporated into the secondary side model in support of the plant training simulator and the analysis of secondary side transients. Results of this analysis are considered preliminary until benchmarked against actual plant data. A model description of the feedwater heater train from the condensate pumps to the deaerator is presented. The methodology used to develop the model is also discussed. Results of the steady state run are presented, and a transient, the loss of extraction steam to feedwater heater 15A, is examined

Feedwater connection repair and modification at GKN

Energy Technology Data Exchange (ETDEWEB)

Witteman, C; Klees, J E

1985-03-01

From January to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing.

Feedwater connection repair and modification at GKN

International Nuclear Information System (INIS)

Witteman, C.; Klees, J.E.

1985-01-01

From Jan. to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing

Getting the most out of your new plant with a chordal ultrasonic feedwater flow measurement system

International Nuclear Information System (INIS)

Estrada, Herb; Hauser, Ernie

2007-01-01

The economic advantages of a chordal ultrasonic feedwater flow measurement system over conventional (flow nozzle-based) feedwater instrumentation are analyzed for new plants having ratings ranging from 1100 MWe to 1600 MWe. Specifically, each of the following topics is considered: The value of a 1.7% increase in the rating of the new plant, made possible by the reduced uncertainty in the determination of thermal power. The value of reduced startup time owing to enhanced steam supply water level control. The value of the reduced feedwater pumping power brought about by the elimination of flow nozzles. The value of the reduced calibration burden owing to the elimination of the feedwater flow differential pressure transmitters and resistance thermometers. The net difference in the acquisition costs of the ultrasonic system versus conventional feedwater flow instrumentation. The net savings in installation costs of the ultrasonic system vis-a-vis conventional feedwater flow instrumentation. The potential savings in outage time due to the reduced frequency of low steam supply water level trips (scrams) of the reactor. (author)

Experimental Results of A1.2 Test for OECD-ATLAS Project

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In order to meet the international interests in the multiple high-risk design extension conditions (DECs) raised after the Fukushima accident, KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In particular, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions in the SBO transients. As one of the new safety improvement concepts to mitigate an SBO accident efficiently, a cooling and operational performance of the passive auxiliary feedwater system (PAFS) is investigated in the framework of the OECD-ATLAS project to produce clearer knowledge of the actual phenomena and to provide the best guidelines for accident management. As the second test of the OECD-ATLAS project, the A1.2 test was conducted to simulate a prolonged SBO with asymmetric secondary cooling through the supply of passive auxiliary feedwater only to SG-2. When the collapsed water level of steam generator reached a wide range of 25%, PAFS was actuated. PAFS played a key role in cooling down the primary system by the heat transfer and the natural circulation. With the actuation of PAFS, the fluid temperatures at the core inlet and outlet started to decrease without any excursion of the maximum heater surface temperature in the core. This integral effect test data of A1.2 test can be used to evaluate the prediction capability of existing safety analysis codes and identify any code deficiency for an SBO simulation with an operation of a passive system such as PAFS.

Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

International Nuclear Information System (INIS)

Tsai, Y.L.; Wang, Li. H.; Fan, T.W.; Ranganath, Sam; Wang, C.K.; Chou, C.P.

2010-01-01

Inspection of the weld between the feedwater nozzle and the safe end at one Taiwan BWR showed axial indications in the Alloy 182 weld. The indication was sufficiently deep that continued operation could not be justified considering the crack growth for one cycle. A weld overlay was decided to implement for restoring the structural margin. This study reviews the cracking cases of feedwater nozzle welds in other nuclear plants, and reports the lesson learned in the engineering project of this weld overlay repair. The overlay design, the FCG calculation and the stress analysis by FEM are presented to confirm that the Code Case structural margins are met. The evaluations of the effect of weld shrinkage on the attached feedwater piping are also included. A number of challenges encountered in the engineering and analysis period are proposed for future study.

Factors analysis of water hammer in FLOWMASTER for main feedwater systems of PWR nuclear power plants

International Nuclear Information System (INIS)

Wang Xin; Han Weishi

2010-01-01

The main feedwater system of a nuclear power plant (NPP) is an important part in ensuring the cooling of a steam generator. It is the main pipe section where water hammers frequently occur. Studying the regulator patterns of water hammers in the main feedwater systems is significant to the stable operation of the system. This article focuses on a parametric study to avoid the consequences of water hammer effect in PWR by employing a general purpose fluid dynamic simulation software-FLOWMASTER. Through FLOWMASTER's transient calculating functions, a mathematical model is established with boundary conditions such as feedwater pumps, control valves, etc., calculations of water hammer pressure when feedwater pumps and control valves shut down, and simulations during instantaneous changes in water hammer pressure. Combining a plethora of engineering practical examples, this research verified the viability of calculating water hammer pressure through FLOWMASTER's transient functions and we found out that, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively. We also found out that changing the intervals of closing signals to feedwater pumps and control valves aid to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (author)

Architectural design of passive solar residential building

Directory of Open Access Journals (Sweden)

Ma Jing

2015-01-01

Full Text Available This paper studies thermal environment of closed balconies that commonly exist in residential buildings, and designs a passive solar residential building. The design optimizes the architectural details of the house and passive utilization of solar energy to provide auxiliary heating for house in winter and cooling in summer. This design might provide a more sufficient and reasonable modification for microclimate in the house.

Impact of the operation of non-displaced feedwater heaters on the performance of Solar Aided Power Generation plants

International Nuclear Information System (INIS)

Qin, Jiyun; Hu, Eric; Nathan, Graham J.

2017-01-01

Highlights: • Impact of non-displaced feedwater heater on plant’s performance has been evaluated. • Two operation strategies for non-displaced feedwater heater has been proposed. • Constant temperature strategy is generally better. • Constant mass flow rate strategy is suit for rich solar thermal input. - Abstract: Solar Aided Power Generation is a technology in which low grade solar thermal energy is used to displace the high grade heat of the extraction steam in a regenerative Rankine cycle power plant for feedwater preheating purpose. The displaced extraction steam can then expand further in the steam turbine to generate power. In such a power plant, using the (concentrated) solar thermal energy to displace the extraction steam to high pressure/temperature feedwater heaters (i.e. displaced feedwater heaters) is the most popular arrangement. Namely the extraction steam to low pressure/temperature feedwater heaters (i.e. non-displaced feedwater heaters) is not displaced by the solar thermal energy. In a Solar Aided Power Generation plants, when solar radiation/input changes, the extraction steam to the displaced feedwater heaters requires to be adjusted according to the solar radiation. However, for the extraction steams to the non-displaced feedwater heaters, it can be either adjusted accordingly following so-called constant temperature strategy or unadjusted i.e. following so-called constant mass flow rate strategy, when solar radiation/input changes. The previous studies overlooked the operation of non-displaced feedwater heaters, which has also impact on the whole plant’s performance. This paper aims to understand/reveal the impact of the two different operation strategies for non-displaced feedwater heaters on the plant’s performance. In this paper, a 300 MW Rankine cycle power plant, in which the extraction steam to high pressure/temperature feedwater heaters is displaced by the solar thermal energy, is used as study case for this purpose. It

Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

International Nuclear Information System (INIS)

Chang, K.C.; Adams, T.M.

1983-01-01

System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

A connection of the steam generator feedwater section of WWER type nuclear power plants

International Nuclear Information System (INIS)

Matal, O.; Sadilek, J.

1989-01-01

In the feedwater piping of each steam generator, a plate for additional water pressure reduction is inserted before the first closing valve. During a steady water flow, the plate gives rise to a constant hydraulic resistance, bringing about steady reduction of the feedwater pressure; this also contributes to a stabilization of the feedwater flow rate into the steam generator. The control valve thus is stressed by minimal hydrodynamic forces. In this manner its load is decreased, its vibrations are damped, and the frequency of failures - and thereby the frequency of the nuclear power plant unit outages -is reduced. (J.P.). 1 fig

Evaluation of examination techniques for ferritic stainless steel feedwater heater tubing

International Nuclear Information System (INIS)

Nugent, M.J.; Catapano, M.C.

1995-01-01

Ferritic stainless steel has been finding increased application in utility plant feedwater heaters due to good strength and corrosion resistance and absence of potential copper contamination of feedwater system. Ferritic stainless steel is highly magnetic and is generally not inspectable using conventional eddy current testing techniques. A variety of techniques have been developed for inspection of this tubing material used in typical heat exchanger applications. Through a project funded by the Empire State Electric Energy Research Corporation (ESEERCO), the evaluation of data generated by four present state of the art NDE testing techniques were evaluated on a controlled mock-up of the heater tubing with service related defects. The primary objective was to determine the strengths and limitations of each method. The testing of two in service feedwater heaters at the Consolidated Edison Company of New York, Inc. (Con Edison's) Arthur Kill Generating Station also allowed further evaluations based on actual field conditions

Feedwater flow measurements: challenges, current solutions, and 'soft' developments

International Nuclear Information System (INIS)

Ruan, D.; Roverso, D.; Fantoni, P.F.; Sanabrias, J.I.; Carrasco, J.A.; Fernandez, L.

2002-07-01

This report presents an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of feedwater flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross correlation flow meter concepts. In this report, Section 2 outlines relevant aspects of thermal power calculations on electrical power plants. Section 3 reviews from the available literature possible approaches and solutions for feedwater flow measurement, including ultrasonic flow meters, cross-correlation flow meters, and 'Virtural' flow meters with artificial neural networks. Section 4 reports typical experimental measurements at the Tecnatom's facility. Section 5 presents an integration approach and preliminary experimental tests. Section 6 discusses the role of soft computing techniques in the context of feedwater flow measurements related nuclear fields, and Section 7 highlights the future research direction. (Author)

Signal validation and failure correction algorithms for PWR steam generator feedwater control

International Nuclear Information System (INIS)

Nasrallah, C.N.; Graham, K.F.

1986-01-01

A critical contributor to the reliability of a nuclear power plant is the reliability of the control systems which maintain plant operating parameters within desired limits. The most difficult system to control in a PWR nuclear power plant and the one which causes the most reactor trips is the control of the feedwater flow to the steam generators. The level in the steam generator must be held within relatively narrow limits, with reactor trips set for both too high and too low a level. The steam generator level is inherently unstable in that it is an open integrator of feedwater flow steam flow mismatch. The steam generator feedwater control system relies on sensed variables in order to generate the appropriate feedwater valve control signal. In current systems, each of these sensed variables comes from a single sensor which may be a separate control sensor or one of the redundant protection sensors that is manually selected by the operator. In case this single signal is false, either due to sensor malfunction or due to a test signal being substituted during periodic test and maintenance, the control system will generate a wrong control signal to the feedwater control valve. This will initiate a steam generator level upset. The solution to this problem is for the control system to sense a given variable with more than one redundant sensor. Normally there are three or four sensors for each variable monitored by the reactor protection system. The techniques discussed allow the control system to compare these redundant sensor signals and generate a validated signal for each measured variable that is insensitive to false signals

Mobile polishing system of feedwater at start-up feedback from the implementation and future prospects

International Nuclear Information System (INIS)

Faure, Celine; Eade, Kevin; Fontan, Guillaume

2012-09-01

offered increased operational flexibility. Its use is limited to the start-up of a unit up to the point where the chemical criteria are met for the changeover from auxiliary feedwater to final feedwater and, at most, up to 20% Pn. In 2006, the will to optimise the discharge during start-up in order to increase the availability of the waste tanks, and to reduce the quantity of metallic oxides in the SGs at the start-up following the appearance of the clogging of SGs broached-hole tube support plates, lead EDF to recreate identical SMEs for the rest of the fleet. Using the SME at the start-up is beneficial for safety, availability, lifespan and discharge reduction. This paper develops the context and technical choices which lead EDF to implement the SME; it goes on to present the design of the treatment system, as well as the implementation feedback of its use on a number of plants. (authors)

Thermal analysis and design of passive solar buildings

CERN Document Server

Athienitis, AK

2013-01-01

Passive solar design techniques are becoming increasingly important in building design. This design reference book takes the building engineer or physicist step-by-step through the thermal analysis and design of passive solar buildings. In particular it emphasises two important topics: the maximum utilization of available solar energy and thermal storage, and the sizing of an appropriate auxiliary heating/cooling system in conjunction with good thermal control.Thermal Analysis and Design of Passive Solar Buildings is an important contribution towards the optimization of buildings as systems th

Simulation of the behaviour of a servo actuated check valve upon rupture of the feedwater pipe

International Nuclear Information System (INIS)

Lucas, A.M. de; Perezagua, R.L.; Rosa, B. de la; Sanz, J.

1995-01-01

The steam generator replacement programme at Almaraz NPP, provides for the installation of a replacement damped non-return valve for the feedwater system. the function of this valve is to protect the steam generator in the event of a rupture in the feedwater pipe. Sudden closure of the check valve, against the flow and following rupture of the feedwater pipe, causes overpressure in the valve which is transmitted to the steam generator nozzle. It is therefore necessary to know this when designing the internal systems of the steam generator. Using the RELAP5/MODE3 code, it has been possible to simulate the dynamic behaviour of a check valve upon rupture of a feedwater pipe postulated outside the containment. The calculation model has been applied to different types of check valve. (Author)

Performance of the prism reactor's passive decay heat removal system

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

Excessive heat removal due to feedwater system malfunction

International Nuclear Information System (INIS)

Beader, D.; Peterlin, G.

1986-01-01

Excessive heat removal transient of the Krsko Nuclear Power Plant, caused by steam generators feedwater system malfunctions was simulated by RELAP5/MOD1 computer code. The results are increase of power and reactor scram caused by high-high steam generator level. (author)

An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

International Nuclear Information System (INIS)

Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

2005-01-01

Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

Velocity Fields Measurement of Natural Circulation Flow inside a Pool Using PIV Technique

International Nuclear Information System (INIS)

Kim, Seok; Kim, Dong Eok; Youn, Young Jung; Euh, Dong Jin; Song, Chul Hwa

2012-01-01

Thermal stratification is encountered in large pool of water increasingly being used as heat sink in new generation of advanced reactors. These large pools at near atmospheric pressure provide a heat sink for heat removal from the reactor or steam generator, and the containment by natural circulation as well as a source of water for core cooling. For examples, the PAFS (passive auxiliary feedwater system) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus), which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by adopting a natural convection mechanism. In a pool, the heat transfer from the PCHX (passive condensation heat exchanger) contributed to increase the pool temperature up to the saturation condition and induce the natural circulation flow of the PCCT (passive condensate cooling tank) pool water. When a heat rod is placed horizontally in a pool of water, the fluid adjacent to the heat rod gets heated up. In the process, its density reduces and by virtue of the buoyancy force, the fluid in this region moves up. After reaching the top free surface, the heated water moves towards the other side wall of the pool along the free surface. Since this heated water is cooling, it goes downward along the wall at the other side wall. Above heater rod, a natural circulation flow is formed. However, there is no flow below heater rod until pool water temperature increases to saturation temperature. In this study, velocity measurement was conducted to reveal a natural circulation flow structure in a small pool using PIV (particle image velocimetry) measurement technique

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

Experimental Results of A1.1 Test for OECD-ATLAS Project

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung-Ho; Bae, Byoung-Uhn; Park, Yu-Sun; Kim, Jong-Rok; Choi, Nam-Hyun; Choi, Ki-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

KAERI (Korea Atomic Energy Research Institute) is operating an OECD/NEA project (hereafter, OECD-ATLAS project) by utilizing a thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). Considering the importance of the SBO scenario and the related accident mitigation measures, a prolonged SBO scenario was selected as the first test subject worthy of investigation in the OECD-ATLAS project as summarized in Table 1. After the Fukushima accident, design extension conditions (DECs) such as an SBO and a total loss of feed water (TLOFW) attracted wide international attention in that such high-risk multiple failure accidents should be revisited from the viewpoint of the reinforcement of the 'defense in depth' concept. In particular, an SBO is one of the most important DECs because a total loss of heat sink can lead to a core melt-down scenario under high pressure without any proper operator action. As for a prolonged SBO transient of the OECD-ATLAS project, two tests, named A1.1 and A1.2, were determined to be performed. In most nuclear power plants (NPPs), a turbine-driven auxiliary feedwater system was designed to remove the decay heat during the early period of an SBO transient. From a conservative point of view, however, it is necessary to investigate the thermal-hydraulic behaviors of the NPP when a turbine-driven auxiliary feedwater supply is not available during the initial period of an SBO transient and moreover a mobile pump-driven auxiliary feedwater supply can only become realized in the later period of the scenario. In particular, asymmetric heat removal characteristic through the supply of auxiliary feedwater only to one steam generator has its own peculiar importance in terms of safety analysis code validation. With an aim of considering these safety importance, in the A1.1 test, a prolonged SBO transient was simulated with two temporal phases: Phase (I) for a conservative SBO transient

Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

International Nuclear Information System (INIS)

Harrell, J.R.; Fuller, R.W.

1996-01-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. The original design and testing requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

Inferential smart sensing for feedwater flowrate in PWRs

International Nuclear Information System (INIS)

Na, M. G.; Hwang, I. J.; Lee, Y. J.

2006-01-01

The feedwater flowrate that is measured by Venturi flow meters in most pressurized water reactors can be over-measured because of the fouling phenomena that make corrosion products accumulate in the Venturi meters. Therefore, in this work, two kinds of methods, a support vector regression method and a fuzzy modeling method, combined with a sequential probability ratio test, are used in order to accurately estimate online the feedwater flowrate, and also to monitor the status of the existing hardware sensors. Also, the data for training the support vector machines and the fuzzy model are selected by using a subtractive clustering scheme to use informative data from among all acquired data. The proposed inferential sensing and monitoring algorithm is verified by using the acquired real plant data of Yonggwang Nuclear Power Plant Unit 3. In the simulations, it was known that the root mean squared error and the relative maximum error are so small and the proposed method early detects the degradation of an existing hardware sensor. (authors)

Corrosion by galvanic coupling on the steam generator auxiliary feedwater pumps at the level of the steam-tight boxes

International Nuclear Information System (INIS)

Dordonat, M.; Huet, M.

1994-01-01

Corrosion by galvanic coupling in steam generator auxiliary pump is coming from electroplated chromium cracks for the rotor steel, and from chemical KANIGEN nickel cracks for the steam-tight boxes black steel. To avoid galvanic coupling between Cr coating and the rotor steel, first an electrolytic Ni coating is done followed by an electrolytic Cr coating. To avoid galvanic coupling between black steel and graphite rings, black steel is replaced by 316L steel. (A.B.). 1 ref., 7 figs

Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network; Estudo de confiabilidade do Sistema Auxiliar de Agua de Alimentacao de uma central nuclear a agua leve por arvore de falhas e rede Bayesiana

Energy Technology Data Exchange (ETDEWEB)

Lava, Deise Diana

2016-10-01

This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

Hydrothermal carbonization (HTC) of wheat straw: influence of feedwater pH prepared by acetic acid and potassium hydroxide.

Science.gov (United States)

Reza, M Toufiq; Rottler, Erwin; Herklotz, Laureen; Wirth, Benjamin

2015-04-01

In this study, influence of feedwater pH (2-12) was studied for hydrothermal carbonization (HTC) of wheat straw at 200 and 260°C. Acetic acid and KOH were used as acidic and basic medium, respectively. Hydrochars were characterized by elemental and fiber analyses, SEM, surface area, pore volume and size, and ATR-FTIR, while HTC process liquids were analyzed by HPLC and GC. Both hydrochar and HTC process liquid qualities vary with feedwater pH. At acidic pH, cellulose and elemental carbon increase in hydrochar, while hemicellulose and pseudo-lignin decrease. Hydrochars produced at pH 2 feedwater has 2.7 times larger surface area than that produced at pH 12. It also has the largest pore volume (1.1 × 10(-1) ml g(-1)) and pore size (20.2 nm). Organic acids were increasing, while sugars were decreasing in case of basic feedwater, however, phenolic compounds were present only at 260°C and their concentrations were increasing in basic feedwater. Copyright © 2015 Elsevier Ltd. All rights reserved.

Mechanics of integrating root causes into PRAs

International Nuclear Information System (INIS)

Bruske, S.Z.; Cadwallader, L.C.; Stepina, P.L.; Vesely, W.E.

1985-01-01

This paper presents a derivation of root cause importance, root cause data for selected components of a pressurized water reactor auxiliary feedwater system, an Accident Sequence Evaluation Program (ASEP) auxiliary feedwater system model, and the results of root cause importance calculations. The methodology shown herein is straightforward and is easily applied to existing probabilistic risk assessments. Root cause importance can greatly benefit the areas of design, maintenance, and inspection. Root cause importance for various components and circumstances can be evaluated

AC-600 passive ECRHR system and its research program

International Nuclear Information System (INIS)

Chen Bingde; Xiao Zejun; Zhou Renmin; Liu Yiyang

1997-01-01

The secondary-side passive emergency core residual heat removal system (ECRHR System) is an important part of AC-600 PWR passive safety system, with which the core decay heat can be removed through nature circulation in primary and secondary system. Since 1991, the program for AC-600 passive ECRHR system has been conducted to investigate its distinct thermal-hydraulic phenomena, heat removal capability, affecting factors, and to develop computer codes. The test facility, designed according to the power/volume simulating law, is a full pressure and temperature operating loop with volume scaling factor of 1/390. It is composed of main loop system, emergence feedwater system, depression system, heat tracing, I and C system and power supply system. A total of sixteen tests is planned in first stage and fifteen of them have been done. The preliminary result analysis showed that the system has efficient heat removal capability in most conditions and some special thermal hydraulic phenomena, for example, flow fluctuation, which has negative impact on system's nature circulation, were identified

Boiler feedwater treatment using reverse osmosis at Suncor OSG

International Nuclear Information System (INIS)

Brown, T.

1997-01-01

The installation of a new 1000 cu m/hr reverse osmosis water treatment system for boiler feedwater at a Suncor plant was discussed. The selection process began in 1993 when Suncor identified a need to increase its boiler feedwater capacity. The company reviewed many options available to increase the treated water capacity. These included: contracting the supply of treated water, adding additional capacity, replacing the entire plant, reverse osmosis, and demineralization. The eventual decision was to build a new 1000 cu m/hr reverse osmosis water treatment plant with the following key components: a Degremont Infilco Ultra Pulsator Clarifier and a Glegg Water Conditioning multimedia filter, Amberpack softeners and reverse osmosis arrays. The reverse osmosis plant was environmentally favourable over an equivalent demineralization plant. A technical comparison was provided between demineralization and reverse osmosis. The system has proven to be successful and economical in meeting the plants needs. 5 figs

Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

International Nuclear Information System (INIS)

Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young; Sang, Seok Yoon

2014-01-01

Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

Energy Technology Data Exchange (ETDEWEB)

Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Sang, Seok Yoon [Engineering and Technical Center, Korea Hydro, Daejeon (Korea, Republic of)

2014-08-15

Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

Digital feedwater and recirculation flow control for GPUN Oyster Creek

International Nuclear Information System (INIS)

Burjorjee, D.; Gan, B.

1992-01-01

This paper describes the digital system for feedwater and recirculation control that GPU Nuclear will be installing at Oyster Creek during its next outage - expected circa December 1992. The replacement was motivated by considerations of reliability and obsolescence - the analog equipment was aging and reaching the end of its useful life. The new system uses Atomic Energy of Canada Ltd.'s software platform running on dual, redundant, industrial-grade 386 computers with opto-isolated field input/output (I/O) accessed through a parallel bus. The feedwater controller controls three main feed regulating valves, two low flow regulating valves, and two block valves. The recirculation controller drives the five scoop positioners of the hydraulic couplers. The system also drives contacts that lock up the actuators on detecting an open circuit in their current loops

Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G., E-mail: sequega@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

2014-10-15

In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

International Nuclear Information System (INIS)

Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G.

2014-10-01

In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

Perfect and Periphrastic Passive Constructions in Danish

DEFF Research Database (Denmark)

Bjerre, Tavs; Bjerre, Anne

2007-01-01

This paper gives an account of the event and argument structure of past participles and the linking between argument structure and valence structure. It further accounts for how participles form perfect and passiv constructions with auxiliaries. We assume that the same participle form is used...... in both types of construction. Our claim is that the valence structure of a past participle is predictable from its semantic type, and that the valence structure predicts with which auciliary a past participle combines in perfect constructions and whether the past participle may occur in passiv...

A novel feedwater system for the RETRAN model of the Palo Verde nuclear generating station

International Nuclear Information System (INIS)

Secker, P.A.; Webb, J.R.

1988-01-01

This paper presents a feedwater system model which supplies realistic boundary conditions to the RETRAN model of a Palo Verde Nuclear Generating Station reactor plant. The RETRAN thermal hydraulic code is used to analyze nuclear reactor system transients through a generalized thermal hydraulic volume/junction network. The feedwater system model is implemented using the control block modeling option available in the RETRAN code. The output of the control block model is coupled to the thermal hydraulic network by a fill junction. A forward Euler integration scheme is used by RETRAN for control block variables. The feedwater system model is formulated to allow implicit integration within the existing code framework. The potential need for small integration time steps is, therefore, alleviated. The model results are compared with test data

The concept of the innovative power reactor

Directory of Open Access Journals (Sweden)

Sang Won Lee

2017-10-01

Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

2014-01-01

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Extensive feedwater quality control and monitoring concept for preventing chemistry-related failures of boiler tubes in a subcritical thermal power plant

International Nuclear Information System (INIS)

Vidojkovic, Sonja; Onjia, Antonije; Matovic, Branko; Grahovac, Nebojsa; Maksimovic, Vesna; Nastasovic, Aleksandra

2013-01-01

Prevention and minimizing corrosion processes on steam generating equipment is highly important in the thermal power industry. The maintenance of feedwater quality at a level corresponding to the standards of technological designing, followed by timely respond to the fluctuation of measured parameters, has a decisive role in corrosion prevention. In this study, the comprehensive chemical control of feedwater quality in 210 MW Thermal Power Plant (TPP) was carried out in order to evaluate its potentiality to assure reliable function of the boiler and discover possible irregularity that might be responsible for frequent boiler tube failures. Sensitive on-line and off-line analytical instruments were used for measuring key and diagnostic parameters considered to be crucial for boiler safety and performances. Obtained results provided evidences for exceeded levels of oxygen, silica, sodium, chloride, sulfate, copper, and conductivity what distinctly demonstrated necessity of feedwater control improvement. Consequently, more effective feedwater quality monitoring concept was recommended. In this paper, the explanation of presumable root causes of corrosive contaminants was given including basic directions for their maintenance in proscribed limits. -- Highlights: • Feedwater quality monitoring practice in a thermal power plant has been evaluated. • The more efficient feedwater quality control have been applied. • Analysis of feedwater quality parameters has been performed. • Exceeded levels of corrosive contaminants were found. • Recommendations for their maintenance at proscribed values were given

Operational challenges to feedwater/steam generator water level control

Energy Technology Data Exchange (ETDEWEB)

Thomas, V.M.; Whaley, S.D.; Federico, P.A. [Westinghouse Electric Company, Cranberry Township, Pennsylvania (United States)

2012-07-01

Feedwater control and turbine control have historically been at the top of the list of contributors to unplanned outages and forced curtailments in the nuclear industry, and they remain so according to recent industry data. Much has been done and is available by way of measures to improve this area and, in spite of much progress, opportunities remain to extend implementation. Toward this end, this paper aims to focus upon feedwater control and provide background on associated characteristics and attributes as a context for identifying the issues which are key challenges that lie at the root of this concern. Primary groupings of these issues will be discussed in order to better define their nature and to establish a basis for a presentation of the range of solutions which have been implemented and remain available to address them. The need for a systems engineering approach, and the role of I&C and field-mounted equipment to application of these solutions will be discussed. (author)

Collector feedwater supply and stability of the power distribution in a pressurized-water reactor

International Nuclear Information System (INIS)

Budnikov, V.I.; Kosolapov, S.V.; Kramerov, A.Ya.

1980-01-01

It is necessary to determine how the collector feedwater supply affects the disposition of the stability limits and the instability period for the power distribution in such a reactor. The main reason for the fluctuations in feedwater flow rate were shown by additional calculations with the general power regulator switched out to be due to instability on the fundamental in the neutron distribution. The power-level fluctuations are due to oscillation of the feed valve in the level regulator, and consequently to oscillations in the feedwater flow rate. If collector feed is to be employed, it is desirable to improve the response of the pressure control system for the separator drum, because under certain emergency conditions there will be a considerable fall in pressure in the separator drum. The deviation from saturation for the water in the separator drum tube is less in the second method than it is in the first, so the cavitation margin in the principal pumps may be reduced somewhat. Calculations show that this reduction will not occur if the time constant of the turbine synchronizer is about 10 sec. Also, the dynamic characteristics of the nuclear power station in these modes of feedwater supply are appreciably influenced by the parameters of the pressure-control system and the water-level control for the separator drum

Analisis Termal High Pressure Feedwater Heater di PLTU PT. XYZ

Directory of Open Access Journals (Sweden)

Maria Ulfa Damayanti

2017-01-01

Full Text Available Abstrak- PT. XYZ mengoperasikan tiga unit Pembangkit Listrik Tenaga Uap (PLTU unit 3, 7 dan 8 berkapasitas 2.030 MegaWatt. Pada PLTU Paiton unit 7 dan 8 terdapat delapan buah feedwater heater yaitu empat buah Low Pressure Water Heater (LPWH, tiga buah High Pressure Water Heater (HPWH, dan sebuah dearator. Pada PLTU Paiton unit 7 dan 8 terdapat kerusakan pada HPWH 6 yang menyebabkan penurunan efisiensi dari siklus secara keseluruhan. Penurunan efisiensi dapat terjadi karena temperatur feedwater sebelum masuk ke boiler terlalu rendah, sehingga kalor yang dibutuhkan oleh boiler untuk memanaskan feedwater meningkat. Oleh karena itu konsumsi batubara akan meningkat dan menyebabkan terjadi kenaikan biaya operasional harian dalam sistem pembangkit. Dari data Divisi Produksi PT. XYZ Unit 7 dan 8 diperoleh spesifikasi HPWH 6, 7, dan 8 dan propertis fluida dalam HPWH 6, 7, dan 8. Data tersebut digunakan sebagai dasar analisis termal yang meliputi performa masing-masing HPH. Tahap selanjutnya dalam analisis termal adalah memvariasikan beban 25%, 50%, 75%, 100%, dan 105%. Tahap terakhir analisis adalah menghitung performa dengan variasi sumbatan (plug 5%, 10%, 15%, dan 20% sesuai dengan variasi beban. Hasil yang didapatkan dari penelitian tugas akhir ini adalah nilai effectiveness tertinggi tercapai pada pembebanan 100% serta menghasilkan pressure drop tertinggi pada pembebanan 105%, nilai effectiveness terbesar serta nilai pressure drop terkecil terjadi pada zona Condensing, serta sumbatan (plugging pada HPH akan menyebabkan penurunan nilai effectiveness dan kenaikan pressure drop sisi tube.

Interim status report on the revision of ASME PTC 12.1 -- closed feedwater heaters

International Nuclear Information System (INIS)

Stellern, J.L.; Hoobler, J.V.; Milton, J.W.; Welch, T.; Kona, C.; Thompson, H.N.; Tsou, J.L.

1993-01-01

The ASME Performance Test Code (PTC) 12.1-1978 for the performance testing of feedwater heaters is being revised extensively and updated. The committee anticipates that the final draft of the proposed Code will be ready for industry review in 1993. This Code revision will greatly enhance the usefulness and cost effectiveness of feedwater heater performance testing. This paper has been prepared to report on the progress of the committee and to disseminate information on the nature of the revision. Included in this paper are some of the notable changes intended for the Code. The most extensive change is the calculation method, which is described in step-by-step detail. An approach is also described for using ultrasonic flow techniques to test individual or split-string feedwater heaters, when flow nozzles are not available. Additionally some educational information on the use and limitations of ultrasonic measurement instrumentation is included. Discussion is also included on the required uncertainty analysis. 3 refs., 2 figs., 2 tabs

Monitoring the performance of Aux. Feedwater Pump using Smart Sensing Model

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-10-15

Many artificial intelligence (AI) techniques equipped with learning systems have recently been proposed to monitor sensors and components in NPPs. Therefore, the objective of this study is the development of an integrity evaluation method for safety critical components such as Aux. feedwater pump, high pressure safety injection (HPSI) pump, etc. using smart sensing models based on AI techniques. In this work, the smart sensing model is developed at first to predict the performance of Aux. feedwater pump by estimating flowrate using group method of data handing (GMDH) method. If the performance prediction is achieved by this feasibility study, the smart sensing model will be applied to development of the integrity evaluation method for safety critical components. Also, the proposed algorithm for the performance prediction is verified by comparison with the simulation data of the MARS code for station blackout (SBO) events. In this study, the smart sensing model for the prediction performance of Aux. feedwater pump has been developed. In order to develop the smart sensing model, the GMDH algorithm is employed. The GMDH algorithm is the way to find a function that can well express a dependent variable from independent variables. This method uses a data structure similar to that of multiple regression models. The proposed GMDH model can accurately predict the performance of Aux.

Monitoring the performance of Aux. Feedwater Pump using Smart Sensing Model

International Nuclear Information System (INIS)

No, Young Gyu; Seong, Poong Hyun

2015-01-01

Many artificial intelligence (AI) techniques equipped with learning systems have recently been proposed to monitor sensors and components in NPPs. Therefore, the objective of this study is the development of an integrity evaluation method for safety critical components such as Aux. feedwater pump, high pressure safety injection (HPSI) pump, etc. using smart sensing models based on AI techniques. In this work, the smart sensing model is developed at first to predict the performance of Aux. feedwater pump by estimating flowrate using group method of data handing (GMDH) method. If the performance prediction is achieved by this feasibility study, the smart sensing model will be applied to development of the integrity evaluation method for safety critical components. Also, the proposed algorithm for the performance prediction is verified by comparison with the simulation data of the MARS code for station blackout (SBO) events. In this study, the smart sensing model for the prediction performance of Aux. feedwater pump has been developed. In order to develop the smart sensing model, the GMDH algorithm is employed. The GMDH algorithm is the way to find a function that can well express a dependent variable from independent variables. This method uses a data structure similar to that of multiple regression models. The proposed GMDH model can accurately predict the performance of Aux

Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

Energy Technology Data Exchange (ETDEWEB)

Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

1997-12-31

This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

International Nuclear Information System (INIS)

Nurkkala, P.; Hoikkanen, J.

1997-01-01

This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. 'grounded' and 'with goose neck'). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.)

Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

Energy Technology Data Exchange (ETDEWEB)

Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland)

1998-12-31

This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

Equipment reliability and life cycle optimization of a nuclear plant feedwater heater

International Nuclear Information System (INIS)

Thomas, Daniel; Coakley, Michael; Catapano, Michael; Svensson, Eric

2006-01-01

Many papers published over the last 25 years have strongly emphasized the need for an ongoing program of inspection and testing with subsequent failure cause analysis of feedwater heaters. Plants must be run more competitively; therefore, Utilities must lower operation and maintenance costs, while optimizing overall plant efficiency and capacity factor. One recognized area that needs to be addressed in accomplishing this goal is the heat cycle. This paper specifically deals with the feedwater heating system. Utility engineers must monitor feedwater heater performance in order to recognize degradation, identify and mitigate failure mechanisms, and prevent in-service failures thereby optimizing availability. Periodic tube plugging without complete analysis of the degraded/failed areas resolves the immediate need for return to service; however, heater life will not be optimized. This paper illustrates a complete life cycle management inspection, testing, and maintenance program implemented at Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data and lack of root cause analysis, justified a program that included: - Removal of previously installed plugs; - Video-probe inspection of failed areas; - Extraction of tube samples for further analysis; - Eddy current testing of selected tubes; - Evaluation of the condition of 'insurance' plugged tubes for return to service; - Hydrostatic testing of selected individual tubes; - Final repair plan based on the results of the above program. This paper concludes that no single method of inspection or testing should solely be relied upon in establishing: - The extent of actual degraded conditions; - The mechanism(s) of failure; - The details of repair to be implemented. Evaluating all data affords the best chance in arresting problems and optimizing feedwater heater life. Problem heaters should be continuously monitored and inspected over time until the facts

Piping reliability improvement through passive seismic supports

International Nuclear Information System (INIS)

Baltus, R.; Rubbers, A.

1999-01-01

The nuclear plants designed in the 1970's were equipped with large quantities of snubbers in auxiliary piping systems. The experience revealed a poor performance of snubbers during periodic inspection, while non-nuclear facility piping survived through strong earthquakes. Consequently, seismic design rules evolved towards more realistic criteria and passive dynamic supports were developed to reduce snubber quantities. These solutions improve the pipe reliability during normal operation while reducing the radiation exposure in a sample line is presented with the impact on pipe stresses compared to the results obtained with passive supports named Limit Stops. (author)

Auxiliary cooling device for power plant

International Nuclear Information System (INIS)

Yamanoi, Kozo.

1996-01-01

An auxiliary cooling sea water pipeline for pumping up cooling sea water, an auxiliary cooling sea water pipeline and a primary side of an auxiliary cooling heat exchanger are connected between a sea water taking vessel and a sea water discharge pit. An auxiliary cooling water pump is connected to an auxiliary water cooling pipeline on the second side of the auxiliary cooling heat exchanger. The auxiliary cooling water pipeline is connected with each of auxiliary equipments of a reactor system and each of auxiliary equipments of the turbine system connected to a turbine auxiliary cooling water pipeline in parallel. During ordinary operation of the reactor, heat exchange for each of the auxiliary equipments of the reactor and heat exchange for each of the equipments of the turbine system are conducted simultaneously. Since most portions of the cooling devices of each of the auxiliary equipments of the reactor system and each of the auxiliary equipments of the turbine system can be used in common, the operation efficiency of the cooling device is improved. In addition, the space for the pipelines and the cost for the equipments can be reduced. (I.N.)

Improvement of Equipment reliability for Auxiliary Feed Water System

Energy Technology Data Exchange (ETDEWEB)

Deok, Lee Sang; Kwan, Lee Yong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

According to AP913 ER) of INPO, Number of the event related to equipment is higher than others like external or human performance. In the top 25 systems, Auxiliary feed water system is the seventh highest among systems. AWFS consists of many component and complex system and Main Function of AFWS is to supply feedwater to the steam generators for the removal of heat from the RCS(Reactor Coolant System) in event the main feedwater system is unavailable following a transient or accident. Reliability of component means how well operate on demands and monitoring is necessary to keep track of condition of component. If component performance is lower than the required value, corrective action for failure mode should be done. The objective of this study is focused to improve of AF pump by adding the tasks of SHR(System Health Report) into the task of system engineer walkdown of PMT(Preventive Maintenance Template). Increasing the reliability of AF pump will contribute to improvement of reliability of AFWS. Based on operating history, there was high vibration of AF pump during performance test. In that case, there were a lot of maintenance works for normal operation of AF pump. Vibration problem related pump can't be detected by tasks of SE walkdown because it's not running during normal operation except for surveillance test. CHR(Component Health Report) of AF pump in AFWS can be made from necessary part which means monitoring and functional failure because problem of Stand-by pump can be covered by conducting monitoring and analysis of functional failure. To improve reliability of AF pump, walkdown of PMT and SHR should be conducted both in accordance with surveillance test frequency. Health of AF pump based on operation history can be verified first and then can find out which parts of pump are weak. Finally, weak part can be managed intensively and failure can be reduced according to SE walkdown. But this work can be risky and burdensome because all parts of

Using risk-informed asset management for feedwater system preventative maintenance optimization

International Nuclear Information System (INIS)

Kee, Ernest; Sun, Alice; Richards, Andrew; Grantom, Rick; Liming, James; Salter, James

2004-01-01

The initial development of a South Texas Project Nuclear Operating Company process for supporting preventative maintenance optimization by applying the Balance-Of-Plant model and Risk-Informed Asset Management alpha-level software applications is presented. Preventative maintenance activities are evaluated in the South Texas Project Risk-Informed Asset Management software while the plant maintains or improves upon high levels of nuclear safety. In the Balance-Of-Plant availability application, the level of detail in the feedwater system is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture by elaborating on the current model at the super-component level of indenture. The enhanced model and modeling techniques are presented. Results of case studies in feedwater system preventative maintenance optimization sing plant-specific data are also presented. (author)

Implementation of a digital feedwater control system at Dresden Nuclear Power Plant, Units 2 and 3: Final report

International Nuclear Information System (INIS)

Zapotocky, A.; Popovic, J.R.; Fournier, R.D.

1988-12-01

This report describes the Digital Feedwater Control System Implementation at the Dresden 2 or 3 Units of the BWR Nuclear Power Plant owned by the Commonwealth Edison Company. The digital system has been operational in Unit 3 since August 1986, and in Unit 2 since April 1987. The Bailey Control's Network 90 based digital control system replaced the obsolete GE/MAC 5000 analog control system in the reactor feedwater control loop as a ''like-for-like'' replacement. Operational experience from the Digital Feedwater Control installations has been good and the system demonstrated better performance than the old analog systems. 14 refs., 15 figs., 17 tabs

The Acquisition of Auxiliary Syntax: A Longitudinal Elicitation Study. Part 1: Auxiliary BE

Science.gov (United States)

Theakston, Anna L.; Rowland, Caroline F.

2009-01-01

Purpose: The question of how and when English-speaking children acquire auxiliaries is the subject of extensive debate. Some researchers posit the existence of innately given Universal Grammar principles to guide acquisition, although some aspects of the auxiliary system must be learned from the input. Others suggest that auxiliaries can be…

Auxiliary buildings

International Nuclear Information System (INIS)

Lakner, I.; Lestyan, E.

1979-01-01

The nuclear power station represents a complicated and a particular industrial project. Consequently, the design of the auxiliary buildings serving the power station (offices, kitchen, refreshment room, workshops, depots, water treatment plant building, boiler houses, etc.) requires more attention than usual. This chapter gives a short survey of the auxiliary buildings already completed and discusses the problems of their design, location and structure. (author)

Identification of BWR feedwater control system using autoregressive integrated model

International Nuclear Information System (INIS)

Kanemoto, Shigeru; Andoh, Yasumasa; Yamamoto, Fumiaki; Idesawa, Masato; Itoh, Kazuo.

1983-01-01

With the view of contributing toward more reliable interpretation of noise behavior under normal operating conditions, which is essential for correct detection and/or diagnosis of incipient anomalies in nuclear power plants by noise analysis technique, studies has been undertaken of the noise behavior in a BWR feedwater control system, with use made of a multivariate autoregressive modeling technique. Noise propagation mechanisms as well as open- and closed-loop responses in the system are identified from noise data by a method in which an autoregressive integrated model is introduced. The closed-loop responses obtained with this method are compared with transient data from an actual test, and confirmed to be reliable in estimating semi-quantitative features. Other analyses performed with this model also yield results that appear most reasonable in their physical characteristics. These results have demonstrated the effectiveness of the noise analyses technique based on the autoregressive integrated model for evaluating and diagnosing the performance of feedwater control systems. (author)

Auxiliary verbs in Dinka

DEFF Research Database (Denmark)

Andersen, Torben

2007-01-01

Dinka, a Western Nilotic language, has a class of auxiliary verbs which is remarkable in the following four respects: (i) It is unusually large, comprising some 20 members; (ii) it is grammatically homogeneous in terms of both morphology and syntax; (iii) most of the auxiliary verbs correspond...... to adverbs in languages like English, while the rest are tense-aspect markers; and (iv) it is possible to combine several auxiliary verbs in a single clause. For some of the auxiliary verbs there are extant full verbs from which they have evolved. To some extent, therefore, it is possible to observe what...

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

International Nuclear Information System (INIS)

Pappx, L.

1994-01-01

After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

Energy Technology Data Exchange (ETDEWEB)

Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

1995-12-31

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

Energy Technology Data Exchange (ETDEWEB)

Logvinov, S.A.; Titov, V.F. [OKB Gidropress (Russian Federation); Notaros, U.; Lenkei, I. [NPP Paks (Hungary)

1995-12-31

A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

Energy Technology Data Exchange (ETDEWEB)

Logvinov, S A; Titov, V F [OKB Gidropress (Russian Federation); Notaros, U; Lenkei, I [NPP Paks (Hungary)

1996-12-31

A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

Energy Technology Data Exchange (ETDEWEB)

Papp, L [Inst. of Material Engineering, Ostrava (Switzerland)

1996-12-31

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

International Nuclear Information System (INIS)

Papp, L.

1995-01-01

After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

Data report of ROSA/LSTF experiment TR-LF-07. Loss-of-feedwater transient with primary feed-and-bleed operation

International Nuclear Information System (INIS)

Takeda, Takeshi

2016-07-01

An experiment TR-LF-07 was conducted on June 23, 1992 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment TR-LF-07 simulated a loss-of-feedwater transient in a pressurized water reactor (PWR) under assumptions of primary feed-and-bleed operation and total failure of auxiliary feedwater system. A safety injection (SI) signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The PZR and hot leg liquid levels recovered due to the HPI actuation in both loops. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The PZR and hot legs became full of liquid again after the ACC actuation. The primary feed-and-bleed operation by use of the PORV, HPI and ACC systems was effective to core cooling because of no core uncovery. The experiment was terminated when the continuous core cooling was confirmed due to the successive coolant injection from the HPI system even after the ACC termination. The obtained data would be useful to study operator actions and procedures in the PWR multiple fault events which behaviors in the PZR affect. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment TR-LF-07. (author)

Multi-unit shutdown due to boiler feedwater chemical excursion

International Nuclear Information System (INIS)

Diebel, M.E.

1991-01-01

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

Numerical analysis on pool boiling using user defined function

Energy Technology Data Exchange (ETDEWEB)

Ryu, Sung Uk; Jeon, Byong Guk; Kim, Seok; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

PAFS (passive auxiliary feedwater system) adopted in the APR+ (Advanced Power Reactor Plus) of Korea is one such application. When PAFS is activated with an actuation signal, steam from the steam generator passes through heat exchanger tubes submerged in a water tank of the PAFS. Outside these heat exchanger tubes, nucleate boiling phenomena appears. In the present work, a numerical study is reported on three-dimensional transient state pool boiling of water having an immersed heat source. The velocity vector fields during the decrease in the water level are numerically investigated in a pool, and the accuracy of the results is checked by comparing the experimental results conducted using the PIV techniques by Kim et al. These numerical results can be used as basic research data for an analysis and prediction of the natural circulation phenomena in the cooling tank of the passive safety system in a nuclear power plant.

Investigations of repetition rate stability of a mode-locked quantum dot semiconductor laser in an auxiliary optical fiber cavity

DEFF Research Database (Denmark)

Breuer, Stefan; Elsässer, Wolfgang; McInerney, J.G.

2010-01-01

We have investigated experimentally the pulse train (mode beating) stability of a monolithic mode-locked multi-section quantum-dot laser with an added passive auxiliary optical fiber cavity. Addition of the weakly coupled (¿ -24dB) cavity reduces the current-induced shift d¿/dI of the principal...

Remote Visual Testing (RVT) for the diagnostic inspection of feedwater heaters

International Nuclear Information System (INIS)

Nugent, M.J.; Pellegrino, B.A.

1993-01-01

Feedwater heaters are an important component in the overall plant heat rate, reliability, availability, performance and maintenance considerations at power stations. The ability to diagnose heater problems in-situ properly can lead to: (1) Preventative plugging of damaged, but unfailed tubes; (2) In-place repair procedures; (3) Incorporation of corrective actions into replacement designs or heater/unit operations. The benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters are briefly reviewed. All Remote Visual Testing (RVT) including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras are discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing are discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections are discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, are detailed. 13 figs

Feedwater heater tube-to-tubesheet connections

International Nuclear Information System (INIS)

Yokell, S.

1993-01-01

This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

Instrument failure detection of flow measurement in the feedwater system of the Paks Nuclear Power Plant, Hungary

International Nuclear Information System (INIS)

Racz, A.

1990-12-01

The applicability of two different methods for early detection of instrument failures of the flow measurement in feedwater systems are investigated. Both methods are based on Kalman filtering technique of stochastic processes. The reliability of the model for description of a feedwater system is checked by comparing calculated values with measured data. Possible instrument failures are simulated in order to show the capability of the proposed procedures. A practical measurement system arrangement is suggested. (author) 10 refs.; 16 figs.; 4 tabs

Report of the Bulletins and Orders Task Force. Volume II. Appendices

International Nuclear Information System (INIS)

1980-01-01

Appendices include: Office of Inspection and Enforcement bulletins; NRR status report on feedwater transients in BWR plants; orders on Babcock and Wilcox Company plants; letters lifting orders; letters issuing auxiliary feedwater system requirements; letter to licensees of all operating reactors, dated October 30, 1979 concerning short-term lessons learned requirements; and letters approving guidelines for preparation of small-break LOCA operating procedures

Ultrasonic pattern recognition study of feedwater nozzle inner radius indication

International Nuclear Information System (INIS)

Yoneyama, H.; Takama, S.; Kishigami, M.; Sasahara, T.; Ando, H.

1983-01-01

A study was made to distinguish defects on feed-water nozzle inner radius from noise echo caused by stainless steel cladding by using ultrasonic pattern recognition method with frequency analysis technique. Experiment has been successfully performed on flat clad plates and nozzle mock-up containing fatigue cracks and the following results which shows the high capability of frequency analysis technique are obtained

Optimization algorithms intended for self-tuning feedwater heater model

International Nuclear Information System (INIS)

Czop, P; Barszcz, T; Bednarz, J

2013-01-01

This work presents a self-tuning feedwater heater model. This work continues the work on first-principle gray-box methodology applied to diagnostics and condition assessment of power plant components. The objective of this work is to review and benchmark the optimization algorithms regarding the time required to achieve the best model fit to operational power plant data. The paper recommends the most effective algorithm to be used in the model adjustment process.

Study on applicability of evaluation model of manpower needs for dismantling of equipments in FUGEN-1. Dismantling process in 3rd/4th feedwater heater room

International Nuclear Information System (INIS)

Shibahara, Yuji; Izumi, Masanori; Nanko, Takashi; Tachibana, Mitsuo; Ishigami, Tsutomu

2010-10-01

Manpower needs for the dismantling process on the dismantling of equipments in FUGEN 3rd/4th feedwater heater room was calculated with the management data evaluation system (PRODIA Code), and it was inspected whether the conventional evaluation model had applicability for FUGEN or not. It was confirmed that the conventional evaluation model for feedwater heater had no applicability. In comparison of the calculated value with the actual data, we found two difference: 1) the calculated value were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was twice larger than that of 4th feedwater heater, though these equipments were almost same weight. It was found that these were brought 1) by the difference in the work descriptions of dismantling between JPDR and FUGEN, and 2) by that in the cutting number between 3rd feedwater heater and 4th one. The manpower needs for the dismantling of both feedwater heaters were calculated with a new calculation equation reflecting the descriptions of dismantling, and it was found that these results showed the good agreement with the actual data. (author)

Corrosion behaviour of a stream generator tube material in simulated steam generator feedwater containing chlorides and sulphates

Energy Technology Data Exchange (ETDEWEB)

Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P.; Yliniemi, K. [VTT Manufacturing Technology, Espoo (Finland); Buddas, T.; Halin, M.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant (Finland)

2002-07-01

The goal of the present work has been to assess the effect of relatively high concentrations of anionic impurities (Cl{sup -}, SO{sub 4}{sup 2-}) on the corrosion behaviour of Ti-stabilised stainless steel SG tubes in simulated steam generator feed-water. The main observations of this work can be summarised as follows: Sulphate ions seem to be more aggressive than chloride ions towards the primary passive film on 08X18H10T stainless steel. The results may indicate that it is more important to have a low concentration of sulphate ions than of chloride ions in secondary side water when the effects of chemical conditions on tube degradation are considered. The presence of chloride ions seems to weaken the detrimental effect of sulphate ions on the stability of oxide films growing on 08X18H10T stainless steel. No localised corrosion features of 08X18H10T stainless steel were detected in the voltammetric and impedance measurements in solutions containing up to 5000 ppb sulphates, chlorides or both of the anions. (authors)

Corrosion behaviour of a stream generator tube material in simulated steam generator feedwater containing chlorides and sulphates

International Nuclear Information System (INIS)

Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P.; Yliniemi, K.; Buddas, T.; Halin, M.; Tompuri, K.

2002-01-01

The goal of the present work has been to assess the effect of relatively high concentrations of anionic impurities (Cl - , SO 4 2- ) on the corrosion behaviour of Ti-stabilised stainless steel SG tubes in simulated steam generator feed-water. The main observations of this work can be summarised as follows: Sulphate ions seem to be more aggressive than chloride ions towards the primary passive film on 08X18H10T stainless steel. The results may indicate that it is more important to have a low concentration of sulphate ions than of chloride ions in secondary side water when the effects of chemical conditions on tube degradation are considered. The presence of chloride ions seems to weaken the detrimental effect of sulphate ions on the stability of oxide films growing on 08X18H10T stainless steel. No localised corrosion features of 08X18H10T stainless steel were detected in the voltammetric and impedance measurements in solutions containing up to 5000 ppb sulphates, chlorides or both of the anions. (authors)

Automatic regulation of the feedwater turbo-pump capacity for the single-turbine 1000 MW NPP unit

International Nuclear Information System (INIS)

Pavlysh, O.N.; Garbuzov, I.P.; Reukov, Yu.N.

1985-01-01

A schematic of the flow regulators (FR) of feedwater turbo-pumps (FTP) for the single-turbine 1000 MW NPP unit is presented. The FR operate in response to feedwoter signals from FTP or in response to FTP rotor rotational speed and control automatic speed governars. The FR automatic regulation ensures limitation of FTP rotor maximum rotational speed at a feedwater flow rate excess equal to 3600 T/h. The transients in the automatic regulation system are considered. Production tests of FTP FR confirmed the FR operation reliability and the right choice of the regulator concept and structure

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (Author)

Proceedings: Power Plant Electric Auxiliary Systems Workshop

International Nuclear Information System (INIS)

1992-06-01

The EPRI Power Plant Electric Auxiliary Systems Workshop, held April 24--25, 1991, in Princeton, New Jersey, brought together utilities, architect/engineers, and equipment suppliers to discuss common problems with power plant auxiliary systems. Workshop participants presented papers on monitoring, identifying, and solving problems with auxiliary systems. Panel discussions focused on improving systems and existing and future plants. The solutions presented to common auxiliary system problems focused on practical ideas that can enhance plant availability, reduce maintenance costs, and simplify the engineering process. The 13 papers in these proceedings include: Tutorials on auxiliary electrical systems and motors; descriptions of evaluations, software development, and new technologies used recently by electric utilities; an analysis of historical performance losses caused by power plant auxiliary systems; innovative design concepts for improving auxiliary system performance in future power plants

Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

International Nuclear Information System (INIS)

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

1986-01-01

A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

Numerical simulation of condensation phase change flow in an inclined tube with bend

Energy Technology Data Exchange (ETDEWEB)

Jo, Jong Chull; Shin, Byung Soo; Do, Kyu Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Yong Kap [Anflux Co., Seoul (Korea, Republic of)

2012-10-15

The new PWR design named APR+ incorporates a passive auxiliary feedwater system (PAFS) as shown in Fig.1. The PAFS consists of two separate divisions. Each division is equipped with one passive condensation heat exchanger (PCHX), isolation or drain or vent valves, check valves, instrumentation and control, and pipes. It is aligned to feed condensed water to its corresponding steam generator (SG). During the PAFS normal operation, steam being produced in the SG secondary side by the residual heat moves up due to buoyancy force and then flows into the PCHX where steam is condensed on the inner surface of the tubes of which the outer surfaces are cooled by the water stored in the passive condensation cooling tank (PCCT). The condensate is passively fed into the SG economizer by gravity. Because the thermal hydraulic characteristics in the PCHT determine the condensation mass rate and the possibility of system instability and water hammer, it is important to understand the condensation phase change flow in the PCHT. This paper presents a numerical simulation of the condensation phase change flow in the PCHX adopted for the APR+ PAFS.

PWR passive plant heat removal assessment: Joint EPRI-CRIEPI advanced LWR studies

International Nuclear Information System (INIS)

1991-03-01

An independent assessment of the capabilities of the PWR passive plant heat removal systems was performed, covering the Passive Residual Heat Removal (PRHR) System, the Passive Safety Injection System (PSIS) and the Passive Containment Cooling System (PCCS) used in a 600 MWe passive plant (e.g., AP600). Additional effort included a review of the test programs which support the design and analysis of the systems, an assessment of the licensability of the plant with regard to heat removal adequacy, and an evaluation of the use of the passive systems with a larger plant. The major conclusions are as follows. The PRHR can remove core decay heat, prevents the pressurizer from filling with water for a loss-of-feedwater transient, and provides safety-grade means for maintaining the reactor coolant system in a safe shutdown condition for the case where the non-safety residual heat removal system becomes unavailable. The PSIS is effective in maintaining the core covered with water for loss-of-coolant accident pipe breaks to eight inches. The PCCS has sufficient heat removal capability to maintain the containment pressure within acceptable limits. The tests performed and planned are adequate to confirm the feasibility of the passive heat removal system designs and to provide a database for verification of the analytical techniques used for the plant evaluations. Each heat removal system can perform in accordance with Regulatory requirements, with the exception that the PRHR system is unable to achieve the required cold shutdown temperature of 200 F within the required 36-hour period. The passive heat removal systems to be used for the 600 MWe plant could be scaled up to a 900 MWe passive plant in a straightforward manner and only minimal, additional confirmatory testing would be required. Sections have been indexed separately for inclusion on the data base

Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

Energy Technology Data Exchange (ETDEWEB)

Tarankov, G.A.; Trunov, N.B.; Titov, V.F. [OKB Gidropress (Russian Federation); Urbansky, V.V. [Rovno NPP (Ukraine); Lenkei, I.; Notarosh, M. [Paks NPP (Hungary)

1995-12-31

Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

Energy Technology Data Exchange (ETDEWEB)

Tarankov, G A; Trunov, N B; Titov, V F [OKB Gidropress (Russian Federation); Urbansky, V V [Rovno NPP (Ukraine); Lenkei, I; Notarosh, M [Paks NPP (Hungary)

1996-12-31

Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

Test Specification of A1-1 Test for OECD-ATLAS Project

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Moon, Sang-Ki; Lee, Seung-Wook; Choi, Ki-Yong; Song, Chul-Hwa

2014-01-01

In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to meet the international interests in the multiple high-risk DECs raised after the Fukushima accident. The proposed test matrix for the OECD-ATLAS project is summarized in Table 1.. In this study, detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG number 2 (SG-2). A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project. A detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG-2 in order to consider an accident mitigation measure. The pre-test analysis using MARS code was performed with an aim of setting up the detailed test procedures for A1-1 test and also gaining the physical insights for a prolonged SBO transient. In the A1-1 test, a prolonged SBO transient will be simulated with two temporal phases: Phase (I) for conservative SBO transient without supply of turbine-driven auxiliary feedwater and Phase (II) for asymmetric cooling via single trained supply of turbine-driven auxiliary feedwater

Auxiliary Deep Generative Models

DEFF Research Database (Denmark)

Maaløe, Lars; Sønderby, Casper Kaae; Sønderby, Søren Kaae

2016-01-01

Deep generative models parameterized by neural networks have recently achieved state-of-the-art performance in unsupervised and semi-supervised learning. We extend deep generative models with auxiliary variables which improves the variational approximation. The auxiliary variables leave...... the generative model unchanged but make the variational distribution more expressive. Inspired by the structure of the auxiliary variable we also propose a model with two stochastic layers and skip connections. Our findings suggest that more expressive and properly specified deep generative models converge...... faster with better results. We show state-of-the-art performance within semi-supervised learning on MNIST (0.96%), SVHN (16.61%) and NORB (9.40%) datasets....

30 CFR 57.6161 - Auxiliary facilities.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary facilities. 57.6161 Section 57.6161...-Underground Only § 57.6161 Auxiliary facilities. (a) Auxiliary facilities used to store explosive material near work places shall be wooden, box-type containers equipped with covers or doors, or facilities...

45 CFR 707.10 - Auxiliary aids.

Science.gov (United States)

2010-10-01

... 45 Public Welfare 3 2010-10-01 2010-10-01 false Auxiliary aids. 707.10 Section 707.10 Public Welfare Regulations Relating to Public Welfare (Continued) COMMISSION ON CIVIL RIGHTS ENFORCEMENT OF... § 707.10 Auxiliary aids. (a) The Agency shall furnish appropriate auxiliary aids where necessary to...

VGB conference 'Chemistry in the power plant 1984' - VGB feedwater conditioning conference

International Nuclear Information System (INIS)

1984-01-01

The conference bears various aspects of feedwater conditioning for power plant cooling systems and steam generators as well as on the analytical assessment of water quality and its translation into operational method approaches. 5 out of the total 14 papers were entered separately in the database. (RB) [de

Life cycle management, design review, and condition assessment of feedwater heaters

Energy Technology Data Exchange (ETDEWEB)

Gammage, D.; Idvorian, N. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

2012-07-01

OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

Life cycle management, design review, and condition assessment of feedwater heaters

International Nuclear Information System (INIS)

Gammage, D.; Idvorian, N.

2012-01-01

OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

Auxiliary mine ventilation manual

International Nuclear Information System (INIS)

Workplace Safety North

2010-01-01

An adequate ventilation system is needed for air quality and handling in a mine and is comprised of many different pieces of equipment for removing contaminated air and supplying fresh air and thereby provide a satisfactory working environment. This manual highlights auxiliary ventilation systems made up of small fans, ducts, tubes, air movers, deflectors and additional air flow controls which distribute fresh air delivered by the primary system to all areas. A review of auxiliary ventilation is provided. Design, operation and management issues are discussed and guidelines are furnished. This manual is limited to underground hard rock operations and does not address directly other, specific auxiliary systems, either in underground coal mines or uranium mines.

Auxiliary mine ventilation manual

Energy Technology Data Exchange (ETDEWEB)

Workplace Safety North

2010-07-01

An adequate ventilation system is needed for air quality and handling in a mine and is comprised of many different pieces of equipment for removing contaminated air and supplying fresh air and thereby provide a satisfactory working environment. This manual highlights auxiliary ventilation systems made up of small fans, ducts, tubes, air movers, deflectors and additional air flow controls which distribute fresh air delivered by the primary system to all areas. A review of auxiliary ventilation is provided. Design, operation and management issues are discussed and guidelines are furnished. This manual is limited to underground hard rock operations and does not address directly other, specific auxiliary systems, either in underground coal mines or uranium mines.

Lessons learned from the SONGS Unit 1 water hammer event

International Nuclear Information System (INIS)

Chiu, C.

1987-01-01

On November 21, 1985, a water hammer event occurred in horizontal feedwater line B at San Onofre Nuclear Generation Site (SONGS) Unit 1. The SONGS Unit 1 is a three-loop pressurized water reactor designed by Westinghouse Electric Corp. The event was initiated by a differential current trip on the bus of auxiliary transformer C. The root cause of the event was a simultaneous failure of five check valves in the feedwater system. Two of them are located downstream of the feedwater pump, and three of them are located further downstream and on the lines to the steam generators. The failure mechanism was determined to be flow-induced vibration, which caused repeated impact between the disk stud and the disk stop. The water hammer occurred in feedwater line B during the refilling of feedwater lines A, B, and C with auxiliary feedwater. The thermal-hydraulic process to initiate the water hammer and the reason that the water hammer only happened in line B have been fully investigated and explained. A root cause analysis after the event was prompted to answer the following two questions: (1) why did these five check valves fail at that time and not in the preceding 15 yr? (2) why did only these five check valves fail? The scope of the root cause analysis involves an investigation of the valve vibration characteristics, plant operation history, and the maintenance history of the valves. The paper answers these two questions, after a brief study of the vibration characteristics of a check valve

BWR feedwater nozzle and control-rod-drive return line nozzle cracking

International Nuclear Information System (INIS)

Anon.

1981-01-01

In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems

Mechanical (turbines and auxiliary equipment)

CERN Document Server

Sherry, A; Cruddace, AE

2013-01-01

Modern Power Station Practice, Volume 3: Mechanical (Turbines and Auxiliary Equipment) focuses on the development of turbines and auxiliary equipment used in power stations in Great Britain. Topics covered include thermodynamics and steam turbine theory; turbine auxiliary systems such as lubrication systems, feed water heating systems, and the condenser and cooling water plants. Miscellaneous station services, and pipework in power plants are also described. This book is comprised of five chapters and begins with an overview of thermodynamics and steam turbine theory, paying particular attenti

Prediction of Heat Removal Capacity of Horizontal Condensation Heat Exchanger submerged in Pool

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Park, Goon-Cherl [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

As representative passive safety systems, there are the passive containment cooling system (PCCS) of ESBWR, the emergency condenser system (ECS) of the SWR-1000, the passive auxiliary feed-water system (PAFS) of the APR+ and etc. During the nuclear power plant accidents, these passive safety systems can cool the nuclear system effectively via the heat transfer through the steam condensation, and then mitigate the accidents. For the optimum design and the safety analysis of the passive safety system, it is essential to predict the heat removal capacity of the heat exchanger well. The heat removal capacity of the horizontal condensation heat exchanger submerged in a pool is determined by a combination of a horizontal in-tube condensation heat transfer and a boiling heat transfer on the horizontal tube. Since most correlations proposed in the previous nuclear engineering field were developed for the vertical tube, there is a certain limit to apply these correlations to the horizontal tube. Therefore, this study developed the heat transfer model for the horizontal Ushaped condensation heat exchanger submerged in a pool to predict well the horizontal in-tube condensation heat transfer, the boiling heat transfer on the horizontal tube and the overall heat removal capacity of the heat exchanger using the best-estimate system analysis code, MARS.

Condensation heat transfer of a feed-water heater and improvement of its performance

International Nuclear Information System (INIS)

Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi

1995-01-01

In this study, a condensation heat transfer model, coupled with a three-dimensional two-phase flow analysis, was developed. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for the steam velocity effect. The model was verified by condensation heat transfer experiments. In the experiments, 112 horizontal, staggered tubes with an outer diameter of 16mm and length of 0.55m were used. The calculated over-all heat transfer coefficients agreed with the data within ±5% under the inlet quality conditions of 13-100%. Based on a three-dimensional two-phase flow analysis, an improved feed-water heater with support plates, which have flow holes between the upper and lower tube bundles, was designed. The total heat exchange capacity of the improved feed-water heater increased about 6%. (author)

Evolution of carbon steel corrosion in feedwater conditions reproduce by the Fortrand loop

International Nuclear Information System (INIS)

Delaunay, Sophie; Bescond, Aurelien; Mansour, Carine; Bretelle, Jean-Luc

2012-09-01

Fouling and tubes support plate blockage of steam generators (SG) are major problems in the secondary circuit of pressurized water reactor (PWR) plants. Corrosion products (CP) responsible of these phenomena are mainly constituted of magnetite. Limit the amount of these CP, generated in the feedwater system and transported to SG, constitutes one way to limit fouling and blockage of SGs. This work requires the understanding of CP behaviour in the feedwater system conditions. A specific experimental circulating water loop, FORTRAND, was built at EDF to follow the formation, the transport and the deposition of iron oxides in representative conditions of the secondary circuit feedwater system. The test section operating at high temperature (up to 250 deg. C) is made in carbon steel and includes three removable segments while all the other parts of the loop are made in stainless steel. First results confirm the formation of iron oxides on carbon steel and stainless steel surface in the conditions of PWR secondary circuits. The surface characterizations show that magnetite is the corrosion product formed on carbon steel and stainless steel at 220 deg. C and goethite is formed at room temperature on stainless steel. The aim of the most recent tests performed in FORTRAND loop was to follow the evolution of corrosion in the feedwater conditions. Tests were performed in one-phase flow conditions at 150 L.h -1 with a linear velocity of 0.82 m/s at 220 deg. C in morpholine/ammonia/hydrazine medium, at pH 25C equal to 9.2. To conduct this study, a removable segment constituted by ten tubes was added to the loop. Several tests were performed to follow the deposit thickness, the iron lost in solution and the oxide morphology with time from two to nine hundred sixty hours. Chemical conditions were controlled and the reproducibility of the results was confirmed by the observation of three tubes at each test. SEM pictures present kinetics with three steps: after the first hours the

Auxiliary reactor for a hydrocarbon reforming system

Science.gov (United States)

Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

2006-01-17

An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

Qualitative and Quantitative Analysis of Organic Impurities in Feedwater of a Heat-Recovery Steam Generator

Science.gov (United States)

Chichirov, A. A.; Chichirova, N. D.; Filimonova, A. A.; Gafiatullina, A. A.

2018-03-01

In recent years, combined-cycle units with heat-recovery steam generators have been constructed and commissioned extensively in the European part of Russia. By the example of the Kazan Cogeneration Power Station no. 3 (TETs-3), an affiliate of JSC TGK-16, the specific problems for most power stations with combined-cycle power units that stem from an elevated content of organic impurities in the feedwater of the heat-recovery steam generator (HRSG) are examined. The HRSG is fed with highly demineralized water in which the content of organic carbon is also standardized. It is assumed that the demineralized water coming from the chemical water treatment department of TETs-3 will be used. Natural water from the Volga River is treated to produce demineralized water. The results of a preliminary analysis of the feedwater demonstrate that certain quality indices, principally, the total organic carbon, are above the standard values. Hence, a comprehensive investigation of the feedwater for organic impurities was performed, which included determination of their structure using IR and UV spectroscopy techniques, potentiometric measurements, and element analysis; determination of physical and chemical properties of organic impurities; and prediction of their behavior in the HRSG. The estimation of the total organic carbon revealed that it exceeded the standard values in all sources of water comprising the feedwater for the HRSG. The extracted impurities were humic substances, namely, a mixture of humic and fulvic acids in a 20 : 80 ratio, respectively. In addition, an analysis was performed of water samples taken at all intermediate stages of water treatment to study the behavior of organic substances in different water treatment processes. An analysis of removal of the humus substances in sections of the water treatment plant yielded the concentration of organic substances on the HRSG condensate. This was from 100 to 150 μg/dm3. Organic impurities in boiler water can induce

30 CFR 75.331 - Auxiliary fans and tubing.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fans and tubing. 75.331 Section 75... HEALTH MANDATORY SAFETY STANDARDS-UNDERGROUND COAL MINES Ventilation § 75.331 Auxiliary fans and tubing. (a) When auxiliary fans and tubing are used for face ventilation, each auxiliary fan shall be— (1...

Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

Energy Technology Data Exchange (ETDEWEB)

Shah, V.N.; Ware, A.G.; Porter, A.M.

1997-03-01

This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

International Nuclear Information System (INIS)

Shah, V.N.; Ware, A.G.; Porter, A.M.

1997-03-01

This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage

The impact of feedwater and condensate return excursions on boiler system component failures

Energy Technology Data Exchange (ETDEWEB)

Esmacher, Mel J. [GE Water and Process Technologies, The Woodlands, TX (United States); Rossi, Anthony [GE Water and Process Technologies, Trevose, PA (United States)

2010-02-15

During boiler operation, the transport of contaminants in boiler feedwater or condensate return via hardness excursions or transport of metal oxides due to corrosion can cause fouling and subsequent tube failure due to under-deposit corrosion or overheating. Case histories are reviewed and suitable corrective actions discussed. (orig.)

Expert system for nuclear power plant feedwater system diagnosis

International Nuclear Information System (INIS)

Meguro, R.; Kinoshita, Y.; Sato, T.; Yokota, Y.; Yokota, M.

1987-01-01

The Expert System for Nuclear Power Plant Feedwater System Diagnosis has been developed to assist maintenance engineers in nuclear power plants. This system adopts the latest process computer TOSBAC G8050 and the expert system developing tool TDES2, and has a large scale knowledge base which consists of the expert knowledge and experience of engineers in many fields. The man-machine system, which has been developed exclusively for diagnosis, improves the man-machine interface and realizes the graphic displays of diagnostic process and path, stores diagnostic results and searches past reference

A study on the shell wall thinning causes identified through experiment, numerical analysis and ultrasonic test of high-pressure feedwater heater

International Nuclear Information System (INIS)

Hwang, Kyeong Mo; Woo, Lee; Jin, Tae Eun; Kim, Kyung Hoon

2008-01-01

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which accelerates as the operation progresses. Several nuclear power plants in Korea have undergone this damage around the impingement baffle - installed downstream of the high-pressure turbine extraction steam line - inside numbers 5A and 5B feedwater heaters. At that point, the extracted steam from the high-pressure turbine consists in the form of two-phase fluid at high temperature, high pressure and high velocity. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of number 5 high-pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the downscaled experimental data in an effort to determine root causes of the shell wall thinning of the high-pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by the actual wall thickness measured by ultrasonic tests. From the comparison of the results for the local velocity profiles and the wall thinning measurements, the local velocity component only in the y-direction flowing vertically to the shell wall, and not in the x- and z-directions, was analogous to the wall thinning data

Benchmarking Simulation of Long Term Station Blackout Events

International Nuclear Information System (INIS)

Kim, Sung Kyum; Lee, John C.; Fynan, Douglas A.; Lee, John C.

2013-01-01

The importance of passive cooling systems has emerged since the SBO events. Turbine-driven auxiliary feedwater (TD-AFW) system is the only passive cooling system for steam generators (SGs) in current PWRs. During SBO events, all alternating current (AC) and direct current (DC) are interrupted and then the water levels of steam generators become high. In this case, turbine blades could be degraded and cannot cool down the SGs anymore. To prevent this kind of degradations, improved TD-AFW system should be installed for current PWRs, especially OPR 1000 plants. A long-term station blackout (LTSBO) scenario based on the improved TD-AFW system has been benchmarked as a reference input file. The following task is a safety analysis in order to find some important parameters causing the peak cladding temperature (PCT) to vary. This task has been initiated with the benchmarked input deck applying to the State-of-the-Art Reactor Consequence Analyses (SOARCA) Report. The point of the improved TD-AFW is to control the water level of the SG by using the auxiliary battery charged by a generator connected with the auxiliary turbine. However, this battery also could be disconnected from the generator. To analyze the uncertainties of the failure of the auxiliary battery, the simulation for the time-dependent failure of the TD-AFW has been performed. In addition to the cases simulated in the paper, some valves (e. g., pressurizer safety valve), available during SBO events in the paper, could be important parameters to assess uncertainties in PCTs estimated. The results for these parameters will be included in a future study in addition to the results for the leakage of the RCP seals. After the simulation of several transient cases, alternating conditional expectation (ACE) algorithm will be used to derive functional relationships between the PCT and several system parameters

Semi-active engine mount design using auxiliary magneto-rheological fluid compliance chamber

Science.gov (United States)

Mansour, H.; Arzanpour, S.; Golnaraghi, M. F.; Parameswaran, A. M.

2011-03-01

Engine mounts are used in the automotive industry to isolate engine and chassis by reducing the noise and vibration imposed from one to the other. This paper describes modelling, simulation and design of a semi-active engine mount that is designed specifically to address the complicated vibration pattern of variable displacement engines (VDE). The ideal isolation for VDE requires the stiffness to be switchable upon cylinder activation/deactivation operating modes. In order to have a modular design, the same hydraulic engine mount components are maintained and a novel auxiliary magneto-rheological (MR) fluid chamber is developed and retrofitted inside the pumping chamber. The new compliance chamber is a controllable pressure regulator, which can effectively alter the dynamic performance of the mount. Switching between different modes happens by turning the electrical current to the MR chamber magnetic coil on and off. A model has been developed for the passive hydraulic mount and then it is extended to include the MR auxiliary chamber as well. A proof-of-concept prototype of the design has been fabricated which validates the mathematical model. The results demonstrate unique capability of the developed semi-active mount to be used for VDE application.

46 CFR 63.25-1 - Small automatic auxiliary boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Small automatic auxiliary boilers. 63.25-1 Section 63.25... AUXILIARY BOILERS Requirements for Specific Types of Automatic Auxiliary Boilers § 63.25-1 Small automatic auxiliary boilers. Small automatic auxiliary boilers defined as having heat-input ratings of 400,000 Btu/hr...

System code improvements for modelling passive safety systems and their validation

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

Developing the optimum boiler water and feedwater treatment for fossil plants

Energy Technology Data Exchange (ETDEWEB)

Dooley, B [Electric Power Research Inst., Palo Alto, California (United States)

1996-12-01

Over the last two years a new set of cycle chemistry guidelines has been developed for each of the treatments used in fossil plants. These revisions have been based on research conducted over the last ten years, much at the international collaborative level. By careful selection and optimization of the boiler water and feedwater treatments, it will be possible to accrue large financial, maintenance, availability and performance improvements. (au) 14 refs.

7 CFR 15b.37 - Auxiliary aids.

Science.gov (United States)

2010-01-01

... 7 Agriculture 1 2010-01-01 2010-01-01 false Auxiliary aids. 15b.37 Section 15b.37 Agriculture... ACTIVITIES RECEIVING FEDERAL FINANCIAL ASSISTANCE Other Aid, Benefits, or Services § 15b.37 Auxiliary aids... appropriate auxiliary aids to persons with impaired sensory, manual, or speaking skills, where necessary to...

30 CFR 57.8529 - Auxiliary fan systems

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fan systems 57.8529 Section 57.8529 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE... Underground Only § 57.8529 Auxiliary fan systems When auxiliary fan systems are used, such systems shall...

LOBI test BT-15/BT-16 pre-test calculation

International Nuclear Information System (INIS)

Holmes, B.J.

1991-03-01

LOBI is a higher pressure, electrically heated integral system facility simulating a KWU 1300 MW PWR scaled 1:712 by volume, although full scale has been maintained in the vertical direction. In order to complete the test programme before the facility is closed in 1991 a number of tests have been performed in tandem, where the test procedures were compatible. BT-15/BT-16 is one such composite test. In their original form both tests BT-15 and BT-16 simulated a loss of main feedwater transient with delayed auxiliary feedwater injection, with the pumps running in BT-15 and tripped in BT-16. The aim of each test was to investigate the loss of primary/secondary heat transfer as the steam generator secondary sides boiled down, and the subsequent recovery of heat transfer as the auxiliary feedwater was tripped on. Due to a re-scheduling of the test programme there was insufficient time to perform sensitivity studies and so only one, base case, calculation is presented. (author)

Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

International Nuclear Information System (INIS)

Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

2014-01-01

The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

International Nuclear Information System (INIS)

Bennert, J.; Becher, L.

1977-01-01

At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

Development of a multi-path ultrasonic flow meter for the application to feedwater flow measurement in nuclear power plants

International Nuclear Information System (INIS)

Jong, J. C.; Ha, J. H.; Kim, Y. H.; Jang, W. H.; Park, K. S.; Park, M. S.; Park, M. H.

2002-01-01

In this work, we propose a method to measure the feedwater flow using multi-path ultrasonic flow meter (UFM). Since the UFM measures a path velocity at which the ultrasonic wave is propagated, the flow profile may be important to convey the path velocity to the velocity averaged over the entire cross section of the flowing medium. The conventional UFM has used the smooth-wall circular pipe model presented by Nikurades. However, this model covers a lower range which is less than 3.2 million while the Reynolds number of the feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we proposed the non-linear correlation model that combines the ratio between the DP output and proposed the non-linear correlation model that combines the ratio between the DP output and UFM output. Experiments were performed using both computer simulation and newly constructed NPPs' test data. The uncertainty analysis result shows that the proposed method has reasonably lower uncertainty than conventional UFM

Investigation into sensitivity of Darlington boiler 2 feedwater flow calibration factor to boiler level control valve configuration

Energy Technology Data Exchange (ETDEWEB)

Coppens, D. [Darlington Nuclear Generating Station, Ontario Power Generation, Bowmanville, Ontario (Canada); Gurevich, Y. [Daystar Technologies Inc., Toronto, Ontario (Canada); Ton, V. [Inspection and Maintenance Services Div., Ontario Power Generation, Ajax, Ontario (Canada); Zobin, D. [AMEC NSS Ltd., Toronto, Ontario (Canada)

2009-07-01

The Ultrasonic Cross-Correlation Flow Meter (USCCFM) has been used for regular feedwater flow calibration at Darlington NGS since the early nineties. Typical measurement repeatability over the duration of a calibration run (normally several weeks long) is within {+-}0.2%. However, it was recently noticed that BO2 calibration factor experienced sudden changes of close to 1%. The paper will describe several different approaches used for identifying the reason for the observed effect. The investigation has revealed that changes in USCCFM readings are due to the complicated geometry of BO2 feedwater piping and that its accuracy can be as high as a fraction of percent if several readings are averaged around the pipe. (author)

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

Ferromagnetic material inspection for feedwater heater and condenser tubes

International Nuclear Information System (INIS)

Anon.

1987-01-01

In recent years, special ferritic stainless steels, such as AL29-4C/sup TM/, Sea-Cure/sup TM/, E-Brite/sup TM/, 439, and similar alloys have been introduced as tube material in condensers, feedwater heaters, moisture separator/reheaters, and other heat exchangers. In addition, carbon steel tubes are widely used in feedwater heaters and heat exchangers in chemical plants. The main problem with the in-service inspection of these ferritic alloys and carbon steel tubes lies in their highly ferromagnetic properties. These properties severely limit the application of the standard eddy current techniques. The effort was undertaken under EPRI sponsorship to develop a reliable technique for in-service inspection of ferromagnetic tubes. The new method combines the measurement of magnetic flux leakage generated around the defects with measurement of total flux in the tube wall. The heart of the inspection system is a special ID probe that magnetizes the tube and generates signals for any tube defect. A permanent record of inspection is provided with a strip-chart or magnetic tape recorder. The laboratory and field evaluation of this new system demonstrated its very good sensitivity to small defects, its reliability, and its ruggedness. Defects as small as 10% external wall loss in heavy wall carbon steel tube were detected. Tubes in the power plant were inspected at a rate of 300-500 tubes per eight-hour shift. The other advantages of this newly developed technique are its simplicity, low cost of instrumentation, easy data interpretation, and full portability

33 CFR 5.47 - Auxiliary ensign.

Science.gov (United States)

2010-07-01

... matching blue Coast Guard Auxiliary emblem is centered. The white slash shall be at a 70 degree angle, rising away from the hoist. (c) The Auxiliary emblem consists of a disk with the shield of the Coat of...

Removal of Iron Oxide Scale from Feed-water in Thermal Power Plant by Using Magnetic Separation

Science.gov (United States)

Nakanishi, Motohiro; Shibatani, Saori; Mishima, Fumihito; Akiyama, Yoko; Nishijima, Shigehiro

2017-09-01

One of the factors of deterioration in thermal power generation efficiency is adhesion of the scale to inner wall in feed-water system. Though thermal power plants have employed All Volatile Treatment (AVT) or Oxygen Treatment (OT) to prevent scale formation, these treatments cannot prevent it completely. In order to remove iron oxide scale, we proposed magnetic separation system using solenoidal superconducting magnet. Magnetic separation efficiency is influenced by component and morphology of scale which changes their property depending on the type of water treatment and temperature. In this study, we estimated component and morphology of iron oxide scale at each equipment in the feed-water system by analyzing simulated scale generated in the pressure vessel at 320 K to 550 K. Based on the results, we considered installation sites of the magnetic separation system.

Device for the analysis of feedwater and condensation samples from power plants

International Nuclear Information System (INIS)

Mostofin, A.A.; Sorokina, N.S.

1978-01-01

An improved version of a device for automatic measurement of the salt and NH 3 contents of feedwater and condensate samples from nuclear power plants is described. Only one sample is required for determining both values. The invention proposes on the one hand to change the dimensions of a throttle opening and on the other to install a second measuring instrument (conductivity measuring instrument). (UWI) [de

'Better feedwater quality through heat exchange equipment renovation'

International Nuclear Information System (INIS)

Pouzenc, C.

2002-01-01

In a fossil-fired or nuclear steam power plant, the water secondary circuit is a critical part of its thermodynamic cycle, as it achieves conditioning, pressurizing and heating of the condensate to match the conditions required at the steam generator inlet. Furthermore, the power plant electrical output and efficiency depend on availability and performances of each component of this secondary circuit from the condenser to the steam generator. Erosion and corrosion phenomena are at the origin of most significant failures in these components and related interconnecting systems. Feedwater chemistry is, together with the selection of materials and optimization of fluid velocities, one of the key levers to protect, as efficiently as possible, the components of the water secondary. (authors)

Water-hammer in the feed-water pipes for PWR steam generators

International Nuclear Information System (INIS)

Gonnet, Bernard; Leroy, Claude; Oullion, Jean; Yazidjian, J.-C.

1979-01-01

PWR boiler water feed pipes have been known for several years to be affected by violent water-hammer during start-ups and operation of the plant. In view of the varying results of corrective design modifications in America and Europe, FRAMATOME undertook an experimental research programme which resulted in the adoption of cruciform tubes on the feed-water distributor as the most reliable solution. Subsequent tests at Fessenheim I confirmed the effectiveness of this device [fr

Implementation of an advanced digital feedwater control system at the Prairie Island nuclear generating station

International Nuclear Information System (INIS)

Paris, R.E.; Gaydos, K.A.; Hill, J.O.; Whitson, S.G.; Wirkkala, R.

1990-05-01

EPRI Project RP2126-4 was a cooperative effort between TVA, EPRI, and Westinghouse which resulted in the demonstration of a prototype of a full range, fully automatic feedwater control system, using fault tolerant digital technology, at the TVA Sequoyah simulator site. That prototype system also included advanced signal validation algorithms and an advanced man-machine interface that used CRT-based soft-control technology. The Westinghouse Advanced Digital Feedwater Control System (ADFCS) upgrade, which contains elements that were part of that prototype system, has since been installed at Northern States Power's Prairie Island Unit 2. This upgrade was very successful due to the use of an advanced control system design and the execution of a well coordinated joint effort between the utility and the supplier. The project experience is documented in this report to help utilities evaluate the technical implications of such a project. The design basis of the Prairie Island ADFCS signal validation for input signal failure fault tolerance is outlined first. Features of the industry-proven system control algorithms are then described. Pre-shipment hardware-in-loop and factory acceptance testing of the Prairie Island system are summarized. Post-shipment site testing, including preoperational and plant startup testing, is also summarized. Plant data from the initial system startup is included. The installation of the Prairie Island ADFCS is described, including both the feedwater control instrumentation and the control board interface. Modification of the plant simulator and operator and I ampersand C personnel training are also discussed. 6 refs., 14 figs., 3 tabs

Changes in feedwater organic matter concentrations based on intake type and pretreatment processes at SWRO facilities, Red Sea, Saudi Arabia

KAUST Repository

Dehwah, Abdullah

2015-03-01

Transparent exopolymer particles (TEP), natural organic matter, and bacterial concentrations in feedwater are important factors that can lead to membrane biofouling in seawater reverse osmosis (SWRO) systems. Two methods for controlling these concentrations in the feedwater prior to pretreatment have been suggested; use of subsurface intake systems or placement of the intake at a greater depth in the sea. These proposed solutions were tested at two SWRO facilities located along the Red Sea of Saudi Arabia. A shallow well intake system was very effective in reducing the algae and bacterial concentrations and somewhat effective in reducing TEP concentrations. An intake placed at a depth of 9. m below the surface was found to have limited impact on improving water quality compared to a surface intake. The algae and bacteria concentration in the feedwater (deep) was lower compared to the surface seawater, but the overall TEP concentration was higher. Bacteria and TEP measurements made in the pretreatment process train in the plant and after the cartridge filters suggest that regrowth of bacteria is occurring within the cartridge filters.

Changes in feedwater organic matter concentrations based on intake type and pretreatment processes at SWRO facilities, Red Sea, Saudi Arabia

KAUST Repository

Dehwah, Abdullah; Li, Sheng; Almashharawi, Samir; Winters, Harvey; Missimer, Thomas M.

2015-01-01

Transparent exopolymer particles (TEP), natural organic matter, and bacterial concentrations in feedwater are important factors that can lead to membrane biofouling in seawater reverse osmosis (SWRO) systems. Two methods for controlling these concentrations in the feedwater prior to pretreatment have been suggested; use of subsurface intake systems or placement of the intake at a greater depth in the sea. These proposed solutions were tested at two SWRO facilities located along the Red Sea of Saudi Arabia. A shallow well intake system was very effective in reducing the algae and bacterial concentrations and somewhat effective in reducing TEP concentrations. An intake placed at a depth of 9. m below the surface was found to have limited impact on improving water quality compared to a surface intake. The algae and bacteria concentration in the feedwater (deep) was lower compared to the surface seawater, but the overall TEP concentration was higher. Bacteria and TEP measurements made in the pretreatment process train in the plant and after the cartridge filters suggest that regrowth of bacteria is occurring within the cartridge filters.

Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator

International Nuclear Information System (INIS)

Sanchez B, A.

2003-01-01

The simulation process of the component systems of the feedwater of a nucleo electric plant is presented, using several models of reduced order that represent the diverse elements that compose the systems like: the heaters of feedwater, the condenser, the feedwater pump, etc. The integration of the same ones in one simulative structure, and the development of a platform that to give the appearance of to be executed in continuous time, it is the objective of the feedwater simulator, as well as of the SUN-RAH simulator, of which is part. The simulator uses models of reduced order that respond to the observed behavior of a nuclear plant of BWR type. Likewise, it is presented a model of a flow controller of feedwater that will be the one in charge of regulating the demand of the system according to the characteristics and criticize restrictions of safety and controllability, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. The integration of these models, the adaptation of the variables and parameters, are presented in a way that the integration with the other ones models of the remaining systems of the plant (reactor, steam lines, turbine, etc.), be direct and coherent with the principles of thermodynamic cycles relative to this type of generation plants. The design of those graphic interfaces and the environment where the simulator works its are part of those developments of this work. The reaches and objectives of the simulator complement the description of the simulator. (Author)

Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

International Nuclear Information System (INIS)

Deghal Cheridi, Amina Lyria; Chaker, Abla; Loubar, Ahcène

2016-01-01

Highlights: • We simulate the behavior of a steam boiler during feed-water line break accident. • To perform accident analysis of the steam boiler, Relap5/Mod3.2 system code is used. • A Relap5 model of the boiler is developed and qualified at the steady state level. • A good agreement between Relap5 results and available experimental data. • The Relap5 model predicts well the main transient features of the boiler. - Abstract: To ensure the operational safety of an industrial water-tube steam boiler it is very important to assess various accident scenarios in real plant working conditions. One of the most challenging scenarios is the loss of feedwater to the steam boiler. In this paper, a simulation of the behavior of an industrial water-tube radiant steam boiler during feedwater line break accident is discussed. The simulation is carried out using the RELAP5 system code. The steam boiler is installed in an Algerian natural gas liquefaction complex. The simulation shows the capabilities of RELAP5 system code in predicting the behavior of the steam boiler at both steady state and transient working conditions. From another side, the behavior of the steam boiler following the accident shows how the control system can successfully mitigate the effects and consequences of such accident and how the evaporator tubes can undergo a severe damage due to an uncontrolled increase of the wall temperature in case of failure of this system.

The Acquisition of Auxiliary Syntax: A Longitudinal Elicitation Study. Part 2: The Modals and Auxiliary DO

Science.gov (United States)

Rowland, Caroline F.; Theakston, Anna L.

2009-01-01

Purpose: The study of auxiliary acquisition is central to work on language development and has attracted theoretical work from both nativist and constructivist approaches. This study is part of a 2-part companion set that represents a unique attempt to trace the development of auxiliary syntax by using a longitudinal elicitation methodology. The…

Experience feedback of an operation event during the experiment of feed-water pump switch

International Nuclear Information System (INIS)

Sun Shuhai; Li Huasheng; Zhang Hao

2012-01-01

In this paper an event is summarized and analyzed, which caused the quit of the high-pressure heaters and the nuclear power rising, during the experiment of the driven feed-water pump switch. The good experience feedback on this event is brought out through gathering related information of domestic nuclear plants. (authors)

Sensitivity Analysis of Core Damage by Loss of Auxiliary Feed Water during the Extended Loss of All AC Power

Energy Technology Data Exchange (ETDEWEB)

Jeon, Woo Jae; Chung, Soon Il; Hwang, Su Hyun; Lee, Kyung Jin; Lee, Byung Chul [FNC Tech., Yongin (Korea, Republic of); Yun, Duk Joo; Lee, Seung Chan [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2015-10-15

In this study, the reactor core damage time for OPR1000 type Nuclear Power Plant (NPP) was analyzed to develop a strategy to handle ELAP and to apply to the EOP. The reactor core damage time in the ELAP condition was calculated according to the time of Auxiliary Feedwater (AFW) loss. Fukushima accident was caused by long hours of Station Black Out (SBO) caused by natural disaster beyond Design Based Accident (DBA) criteria. It led to the reactor core damage. After the accident, the regulatory authorities of each country (Japan, US, EU, IAEA, and etc.) recommended developing the necessary systems and strategies in order to cover up the Extended Loss of All AC Power (ELAP) such as one occurred in the Fukushima accident. And the need of procedure or guideline to cope with ELAP has been raised through the stress test for Wolsong Nuclear Power Plant unit 1. Current Emergency Operating Procedures (EOP) used in domestic nuclear power plant are seemed to be insufficient to cope with ELAP. Therefore, it has been required to be improved. As the result, the time of AFW loss in the ELAP condition influences greatly on core damage time.

CAREM-25. Auxiliary systems

International Nuclear Information System (INIS)

Acosta, Eduardo; Amaya, Daniel; Carlevaris, Rodolfo; Patrignani, A.; Ramilo, L.; Santecchia, A.; Vindrola, C.

2000-01-01

CAREM is an innovative PWR reactor whose prototype will be of small power generation capacity (100MWt, about 25MWe).CAREM design is based on light water integrated reactor with slightly enriched uranium.In this work, a summary of the functions and most relevant design characteristics of main auxiliary systems associated to the chain of heat removal and physicochemical - radiological treatment of the cooling fluids of the CAREM-25 prototype is presented.Even though these auxiliary systems of the reactor are not safety system, they fulfill functions related with the nuclear safety at different operative modes of the reactor

CAREM-25. Auxiliary systems

International Nuclear Information System (INIS)

Acosta, Eduardo; Amaya, Daniel; Carlevaris, Rodolfo; Patrignani, Alberto; Santecchia, Alberto; Vindrola, Carlos; Ramilo, Lucia B.

2000-01-01

CAREM is an innovative PWR reactor whose prototype will be of small power generation capacity (100 M Wt, about 25 M We). CAREM design is based on light water integrated reactor with slightly enriched uranium. In this work, a summary of the functions and most relevant design characteristics of main auxiliary systems associated to the chain of heat removal and physicochemical - radiological treatment of the cooling fluids of the CAREM-25 prototype is presented. Even though these auxiliary systems of the reactor are not safety system, they fulfill functions related with the nuclear safety at different operative modes of the reactor. (author)

77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Science.gov (United States)

2012-03-16

... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system

International Nuclear Information System (INIS)

Miron, A.

1998-01-01

In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able to reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal

European passive plant program A design for the 21st century

International Nuclear Information System (INIS)

Adomaitis, D.; Oyarzabal, M.

1998-01-01

In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The following major tasks were accomplished: (1) the impacts of the European utility requirements (EUR) on the Westinghouse nuclear island design were evaluated; and (2) a 1000 MWe passive plant reference design (EP1000) was established which conforms to the EUR and is expected to be licensable in Europe. With respect to safety systems and containment, the reference plant design closely follows that of the Westinghouse simplified pressurized water reactor (SPWR) design, while the AP600 plant design has been taken as the basis for the EP1000 reference design in the auxiliary system design areas. However, the EP1000 design also includes features required to meet the EUR, as well as key European licensing requirements. (orig.)

Computation within the auxiliary field approach

International Nuclear Information System (INIS)

Baeurle, S.A.

2003-01-01

Recently, the classical auxiliary field methodology has been developed as a new simulation technique for performing calculations within the framework of classical statistical mechanics. Since the approach suffers from a sign problem, a judicious choice of the sampling algorithm, allowing a fast statistical convergence and an efficient generation of field configurations, is of fundamental importance for a successful simulation. In this paper we focus on the computational aspects of this simulation methodology. We introduce two different types of algorithms, the single-move auxiliary field Metropolis Monte Carlo algorithm and two new classes of force-based algorithms, which enable multiple-move propagation. In addition, to further optimize the sampling, we describe a preconditioning scheme, which permits to treat each field degree of freedom individually with regard to the evolution through the auxiliary field configuration space. Finally, we demonstrate the validity and assess the competitiveness of these algorithms on a representative practical example. We believe that they may also provide an interesting possibility for enhancing the computational efficiency of other auxiliary field methodologies

Bayesian Analysis of Geostatistical Models With an Auxiliary Lattice

KAUST Repository

Park, Jincheol; Liang, Faming

2012-01-01

of observations is large. In this article, we propose an auxiliary lattice-based approach for tackling this difficulty. By introducing an auxiliary lattice to the space of observations and defining a Gaussian Markov random field on the auxiliary lattice, our model

47 CFR 73.1675 - Auxiliary antennas.

Science.gov (United States)

2010-10-01

... Class A TV licensees may request a decrease from the authorized facility's ERP in the license application. An FM, TV or Class A TV licensee may also increase the ERP of the auxiliary facility in a license... licensed main facility as an auxiliary facility with an ERP less than or equal to the ERP specified on the...

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1988-01-01

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

46 CFR 58.25-10 - Main and auxiliary steering gear.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Main and auxiliary steering gear. 58.25-10 Section 58.25... AUXILIARY MACHINERY AND RELATED SYSTEMS Steering Gear § 58.25-10 Main and auxiliary steering gear. (a) Power-operated main and auxiliary steering gear must be separate systems that are independent throughout their...

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

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

1997-01-01

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

Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

International Nuclear Information System (INIS)

Finicle, D.P.

1977-01-01

The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

Power-feedwater enthalpy operating domain for SBWR applying Monte Carlo simulation

International Nuclear Information System (INIS)

Quezada-Garcia, S.; Espinosa-Martinez, E.-G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

2014-01-01

In this work the analyses of the feedwater enthalpy effects on reactor power in a simplified boiling water reactor (SBWR) applying a methodology based on Monte Carlo's simulation (MCS), is presented. The MCS methodology was applied systematically to establish operating domain, due that the SBWR are not yet in operation, the analysis of the nuclear and thermalhydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. (author)

Study on thermal-hydraulic phenomena identification of passive heat removal facilities

International Nuclear Information System (INIS)

Park, J. Y.

2011-01-01

Recently, passive heat removal facilities have been integral features of new generation or future reactor designs worldwide. This is because the passive heat removal facilities depending on a natural force such as buoyancy can give much higher operational reliability compared to active heat removal facilities depending on pumped fluid flow and as a result they can decrease core damage frequency of a nuclear power plant drastically ever achievable before. Keeping pace with this global trend, SMART and APR+ reactors also have introduced passive heat removal features such as a passive residual heat removal system (PRHRS) and a passive auxiliary feed water system (PAFS) in their designs. Since many thermal-hydraulic (T-H) phenomena including steam condensation are involved during operation of the passive heat removal facilities, they ought to be properly simulated by T-H codes such as MARS-KS and RELAP5 in order to guarantee reliable safety analysis by these codes. Unfortunately, however, these T-H codes are not well validated with respect to phenomena related to passive heat removal mechanism because previous focus on these codes validation was mainly on the LB LOCA and resulting phenomena. To resolve this gap, Korea Institute of Nuclear Safety has initiated a research program on the development of safety analysis technology for passive heat removal facilities. The main target of this program is PRHRS and PAFS in SMART and APR+ reactors and through this program, validation of capability of existing T-H codes and improvement of codes regarding passive facilities analysis are to be sought. In part of this research, T-H phenomena important to passive heat removal facilities (PRHRS and PAFS) are investigated in the present study

14 CFR 29.757 - Hull and auxiliary float strength.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Hull and auxiliary float strength. 29.757... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Design and Construction Floats and Hulls § 29.757 Hull and auxiliary float strength. The hull, and auxiliary floats if used, must withstand the...

78 FR 27321 - Revision of Auxiliary Regulations

Science.gov (United States)

2013-05-10

... Auxiliary organizational structure, extending civil liability protection to Auxiliary units and members, and... entities. C. Assistance for Small Entities If you think that your business, organization, or governmental..., explain why you think your business or organization qualifies, how and to what degree this rule would...

ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

International Nuclear Information System (INIS)

Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

1999-01-01

ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

Loss of preferred power events in German BWRs and PWRs

International Nuclear Information System (INIS)

Frisch, W.

1984-01-01

This chapter examines the variety of courses which the loss of auxiliary power case may take in different plants, depending on the initiation of the event, the switching to the alternate auxiliary power supply and the automatic start-up sequences of auxiliaries (e.g. feedwater pumps, condensate pumps) after a short term interruption of auxiliary power. Topics considered include main station transformer failure in a boiling water reactor (BWR), main plant features, event description, evaluation of the event, deviations from expected behavior, operator actions, a comparison with start-up tests and other events, a comparison with postcalculations, and auxiliary transformer failure in a pressurized water reactor (PWR). Loss of preferred (or auxiliary) power events are standard design cases taken into account in the design of many reactor and plant systems

Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

International Nuclear Information System (INIS)

Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

1995-01-01

Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system

Application of a Long Term Asset Management Strategy for HP Feedwater Heaters

International Nuclear Information System (INIS)

Won, Se Youl; Yun, Eun Sub; Park, Young Sheop

2008-01-01

As the commercial operating year of nuclear power plants is increased, it becomes imperative to develop integrated cost-effective asset management and to improve plans for degraded Structures, Systems, and Components (SSCs) in terms of safety and economical consideration. A long-term asset management (LTAM) strategy can improve the condition of nuclear plants, maximize their value, and optimize their operational life by maintaining their safety. This paper presents an optimized LTAM plan for HP feedwater heaters at a specific nuclear power plant

Vacuum switchgear for power station auxiliary switchboards

International Nuclear Information System (INIS)

Coombs, P.E.

1992-01-01

Sizewell B is the first UK power station in which vacuum switchgear is used for the auxiliary switchboards. Previously the 3.3kV, 6.6kV or 11kV switchgear has used air-break circuit breakers and fused air-break contactors, known as motor starting devices or fused switching devices (FSD). The use of vacuum interrupters is therefore a new technology in this application, although it has been established in the UK distribution network and in industrial installations from the mid 1970s. Vacuum switchgear was already in use in the USA for power station auxiliary switchgear at the time that it was proposed for Sizewell B. The Sizewell B high voltage auxiliary switchgear comprises eight Unit and Station Auxiliary Switchboards at 3.3kV and 11kV, and four 3.3kV Essential Switchboards for the essential safety related circuits, making a total of 65 circuit breakers plus FSD panels. (Author)

30 CFR 57.22209 - Auxiliary fans (I-C mines).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fans (I-C mines). 57.22209 Section 57... Standards for Methane in Metal and Nonmetal Mines Ventilation § 57.22209 Auxiliary fans (I-C mines). Electric auxiliary fans shall be approved by MSHA under the applicable requirements of 30 CFR part 18...

Improving Energy Efficiency of Auxiliaries

International Nuclear Information System (INIS)

Carl T. Vuk

2001-01-01

The summaries of this report are: Economics Ultimately Dictates Direction; Electric Auxiliaries Provide Solid Benefits. The Impact on Vehicle Architecture Will be Important; Integrated Generators With Combined With Turbo Generators Can Meet the Electrical Demands of Electric Auxiliaries; Implementation Will Follow Automotive 42V Transition; Availability of Low Cost Hardware Will Slow Implementation; Industry Leadership and Cooperation Needed; Standards and Safety Protocols Will be Important. Government Can Play an Important Role in Expediting: Funding Technical Development; Incentives for Improving Fuel Economy; Developing Standards, Allowing Economy of Scale; and Providing Safety Guidelines

MARS-KS Code Assessment for Condensation Heat Transfer in Horizontal Tube with the Presence of Non-Condensable Gas using Purdue Experiment

International Nuclear Information System (INIS)

Jeon, Seong Su; Lee, Byung Chul; Park, Ju Yeop; Seul, Kwang Won

2011-01-01

In South Korea, advanced power reactor plus (APR+), as a Korean specific reactor, is currently under development for the export strategy. In order to raise competitiveness of the APR+ in the world market, it is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a passive auxiliary feedwater system (PAFS) was adopted as an improved safety design concept of APR+: and then there have been many efforts to develop the PAFS. According to PAFS design concept, PAFS can completely replace the auxiliary feedwater system. When the design basis accident, in which feedwater is unavailable, occurs, the PAFS can remove the residual heat in the core and then prevent the core damage. In the PAFS with the horizontal type heat exchanger, two-phase natural circulation, condensation heat transfer in tube, boiling heat transfer in pool, natural convection in pool, etc. are considered as very important thermalhydraulic phenomena (see Fig. 1). Compared with the vertical heat exchanger from these phenomena, the major difference of the horizontal heat exchanger is the condensation heat transfer phenomena in the tube side. There have been many efforts to understand the condensation heat transfer with in the presence of NC gas in tube but most researches focused on the condensation heat transfer in vertical tube. Therefore the details of the condensation heat transfer in the presence of NC gas in horizontal condenser tubes are not well understood. In order to develop the safety evaluation system for APR+ PAFS, it is required to evaluate the capability and applicability of the MARS-KS code for modeling the condensation heat transfer in the horizontal tube with NC gas because many heat transfer correlations in MARS-KS are known to have much uncertainty. In particular, there is no reliable model for the condensation phenomena in horizontal tube with NC gas. In order to assess the MARS-KS code results and identify the

Analysis of KNU1 loss of normal feedwater

International Nuclear Information System (INIS)

Kim, Hho-Jung; Chung, Bub-Dong; Lee, Young-Jin; Kim, Jin-Soo

1986-01-01

Simulation of the system thermal-hydraulic parameters was carried out following the KNU1 (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant system) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (author)

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

International Nuclear Information System (INIS)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

2016-01-01

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

Energy Technology Data Exchange (ETDEWEB)

Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

A reliability assessment methodology for the VHTR passive safety system

International Nuclear Information System (INIS)

Lee, Hyungsuk; Jae, Moosung

2014-01-01

The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

The installation of helium auxiliary systems in HTGR

International Nuclear Information System (INIS)

Qin Zhenya; Fu Xiaodong

1993-01-01

The inert gas Helium was chosen as reactor coolant in high temperature gas coolant reactor, therefore a set of Special and uncomplex helium auxiliary systems will be installed, the safe operation of HTR-10 can be safeguarded. It does not effect the inherent safety of HTR-10 MW if any one of all those systems were damaged during operation condition. This article introduces the design function and the system principle of all helium auxiliary systems to be installed in HTR-10. Those systems include: helium purification and its regeneration system, helium supply and storage system, pressure control and release system of primary system, dump system for helium auxiliary system and fuel handling, gaseous waste storage system, water extraction system for helium auxiliary systems and evacuation system for primary system

Remote visual testing (RVT) for the diagnostic inspection of feedwater heaters

International Nuclear Information System (INIS)

Nugent, M.J.; Pellegrino, B.A.

1991-01-01

In this paper the benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters will be briefly reviewed. All Remote Visual Testing (RVT) devices including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras will be discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing will be discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections will be discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, will be detailed

Auxiliary resonant DC tank converter

Science.gov (United States)

Peng, Fang Z.

2000-01-01

An auxiliary resonant dc tank (ARDCT) converter is provided for achieving soft-switching in a power converter. An ARDCT circuit is coupled directly across a dc bus to the inverter to generate a resonant dc bus voltage, including upper and lower resonant capacitors connected in series as a resonant leg, first and second dc tank capacitors connected in series as a tank leg, and an auxiliary resonant circuit comprising a series combination of a resonant inductor and a pair of auxiliary switching devices. The ARDCT circuit further includes first clamping means for holding the resonant dc bus voltage to the dc tank voltage of the tank leg, and second clamping means for clamping the resonant dc bus voltage to zero during a resonant period. The ARDCT circuit resonantly brings the dc bus voltage to zero in order to provide a zero-voltage switching opportunity for the inverter, then quickly rebounds the dc bus voltage back to the dc tank voltage after the inverter changes state. The auxiliary switching devices are turned on and off under zero-current conditions. The ARDCT circuit only absorbs ripples of the inverter dc bus current, thus having less current stress. In addition, since the ARDCT circuit is coupled in parallel with the dc power supply and the inverter for merely assisting soft-switching of the inverter without participating in real dc power transmission and power conversion, malfunction and failure of the tank circuit will not affect the functional operation of the inverter; thus a highly reliable converter system is expected.

Plant data comparisons for Comanche Peak 1/2 main feedwater pump trip transient

Energy Technology Data Exchange (ETDEWEB)

Boatwright, W.J.; Choe, W.G; Hiltbrand, D.W. [TU Electric, Dallas, TX (United States)] [and others

1995-09-01

A RETRAN-02 MOD5 model of Comanche Peak Steam Electric Station was developed by TU Electric for the purpose of performing core reload safety analyses. In order to qualify this model, comparisons against plant transient data from a partial loss of main feedwater flow were performed. These comparisons demonstrated that good representations of the plant response could be obtained with RETRAN-02 and the user-developed models of the primary-to-secondary heat transfer and plant control systems.

Comparative Study of the Passive Verb in Arabic and Persian Languages from the Perspective of Grammatical and Semantic

Directory of Open Access Journals (Sweden)

Mansooreh Zarkoob

2012-11-01

Full Text Available Abstract Verb is one the important categories and main elements of sentence which is sometimes divided in similar types in Arabic and Persian. One of the main types of verb existed in both languages is passive verb. Although this appellation is apparently common in both languages, it seems passive verbs are completely equivalent in both languages but since passive verb in the Persian language has been discussed from different aspects compared with Arabic, in this article we are looking for some answers to these questions that if we can find other structures apart from passive structure which are accounted as passive in their meanings? Which kind of relationship is there between grammatical and semantic structures of passive verb in both languages? What are grammatical and semantic differences and similarities of passive verb in Persian and Arabic languages? The results of this survey decreased translation errors of students. We also state this example as a result of this research that, not only there is an auxiliary verb in both languages is investigated as a passive-maker but also, there are some planar verbs in both languages and also voice changes are occurred in addition to inflection changes.

Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II Nuclear Power Plant

International Nuclear Information System (INIS)

Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

1993-12-01

The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and, a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation. This transient actually occurred in the plant on June 19, 1989

Evaluation of heatup and recovery in a loss of feedwater accident with multiple failure

International Nuclear Information System (INIS)

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

1991-01-01

A loss of feedwater accident with multiple failure has been studied in order to identify the potential severity of the accident when compared with the design basis accident in PWR. The PCS heatup and recovery mode in a LOFA with multiple failure was evaluated using the LOFT L9-1/L3-3 experiment. From experimental result, 4 separable subphase were identified and the associated phenomena were also addressed

Tracer test method and process data reconciliation based on VDI 2048. Comparison of two methods for highly accurate determination of feedwater massflow at NPP Beznau

International Nuclear Information System (INIS)

Hungerbuehler, T.; Langenstein, M.

2007-01-01

The feedwater mass flow is the key measured variable used to determine the thermal reactor output in a nuclear power plant. Usually this parameter is recorded via venturi nozzles of orifice plates. The problem with both principles of measurement, however, is that an accuracy of below 1% cannot be reached. In order to make more accurate statements about the feedwater amounts recirculated in the water-steam cycle, tracer measurements that offer an accuracy of up to 0.2% are used. In the NPP Beznau both methods have been used in parallel to determine the feedwater flow rates in 2004 (unit 1) and 2005 (unit 2). Comparison of the results shows that a high level of agreement is obtained between the results of the reconciliation and the results of the tracer measurements. As a result of the findings of this comparison, a high level of acceptance of process data reconciliation based on VDI 2048 was achieved. (orig.)

Excitation of waves in plasma near the ion cyclotron frequency using surface-wave antennas with auxillary passive gaps

International Nuclear Information System (INIS)

Longinov, A.V.; Lukinov, V.A.

1992-01-01

It is proposed to use a system of auxiliary passive gaps to excite waves in a plasma traveling in one direction parallel to the magnetic field, in order to localize the radiating surface of a surface-wave antenna. Using excitation of ion Bernstein waves in the plasma as an example the main properties of such an antenna system have been studied. It is shown that the use of passive gaps permits high directionality to be achieved for the radiation and allows the size of the radiating surface of the antenna to be controlled. 10 refs., 6 figs

Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

Energy Technology Data Exchange (ETDEWEB)

Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

1988-12-31

A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

47 CFR 74.601 - Classes of TV broadcast auxiliary stations.

Science.gov (United States)

2010-10-01

... 47 Telecommunication 4 2010-10-01 2010-10-01 false Classes of TV broadcast auxiliary stations. 74... Television Broadcast Auxiliary Stations § 74.601 Classes of TV broadcast auxiliary stations. (a) TV pickup stations. A land mobile station used for the transmission of TV program material and related communications...

Effects of Auxiliary-Source Connection in Multichip Power Module

DEFF Research Database (Denmark)

Li, Helong; Munk-Nielsen, Stig; Wang, Xiongfei

2017-01-01

the power loop and the gate loop like how the Kelvin-source connection does, owing to their involvement in the loop of the power source current. Three effects of the auxiliary-source connections are then analyzed, which are 1) the common source stray inductance reduction, 2) the transient drain......Auxiliary-source bond wires and connections are widely used in power modules with paralleled MOSFETs or IGBTs. This paper investigates the operation mechanism of the auxiliary-source connections in multichip power modules. It reveals that the auxiliary-source connections cannot fully decouple......-source current imbalance mitigation, and 3) the influence on the steady-state current distribution. Lastly, simulations and experimental results validate the theoretical analysis....

Evaluation method for two-phase flow and heat transfer in a feed-water heater

International Nuclear Information System (INIS)

Takamori, Kazuhide; Minato, Akihiko

1993-01-01

A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

30 CFR 57.8534 - Shutdown or failure of auxiliary fans.

Science.gov (United States)

2010-07-01

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

Auxiliary System Load Schemes in Large Thermal and Nuclear Power Plants

International Nuclear Information System (INIS)

Kuzle, I.; Bosnjak, D.; Pandzic, H.

2010-01-01

Uninterrupted auxiliary system power supply in large power plants is a key factor for normal operation, transient states, start-ups and shutdowns and particularly during fault conditions. Therefore, there are many challenges in designing the main electrical system as well as the auxiliary systems power supply. Depending upon the type of fuel used and the environmental control system required, a thermal power plant may consume as much as 10% of its total generation for auxiliary power, while a nuclear power plant may require only 4 - 6% auxiliaries. In general, the larger the power generating plant, the higher the voltage selected for the AC auxiliary electric system. Most stations in the 75 to 500 MW range utilize 4,2 kV as the base auxiliary system voltage. Large generating stations 500 - 1000 MW and more use voltage levels of 6,9 kV and more. Some single dedicated loads such as electric driven boiler feed pumps are supplied ba a 13,8 kV bus. While designing the auxiliary electric system, the following areas must be considered: motor starting requirements, voltage regulation requirements, short-circuit duty requirements, economic considerations, reliability and alternate sources. Auxiliary power supply can't be completely generalized and each situation should be studied on its own merits to determine the optimal solution. Naturally, nuclear power plants have more reliability requirements and safety design criteria. Main coolant-pump power supply and continuity of service to other vital loads deserve special attention. This paper presents an overview of some up-to-date power plant auxiliary load system concepts. The main types of auxiliary loads are described and the electric diagrams of the modern auxiliary system supply concepts are given. Various alternative sources of auxiliary electrical supply are considered, the advantages and disadvantages of these are compared and proposals are made for high voltage distribution systems around the thermal and nuclear plant

Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

Builtin vs. auxiliary detection of extrapolation risk.

Energy Technology Data Exchange (ETDEWEB)

Munson, Miles Arthur; Kegelmeyer, W. Philip,

2013-02-01

A key assumption in supervised machine learning is that future data will be similar to historical data. This assumption is often false in real world applications, and as a result, prediction models often return predictions that are extrapolations. We compare four approaches to estimating extrapolation risk for machine learning predictions. Two builtin methods use information available from the classification model to decide if the model would be extrapolating for an input data point. The other two build auxiliary models to supplement the classification model and explicitly model extrapolation risk. Experiments with synthetic and real data sets show that the auxiliary models are more reliable risk detectors. To best safeguard against extrapolating predictions, however, we recommend combining builtin and auxiliary diagnostics.

Common cause evaluations in applied risk analysis of nuclear power plants

International Nuclear Information System (INIS)

Taniguchi, T.; Ligon, D.; Stamatelatos, M.

1983-04-01

Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system

14 CFR 25.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 25.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Powerplant Controls and Accessories § 25.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

14 CFR 23.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 23.1142... Powerplant Controls and Accessories § 23.1142 Auxiliary power unit controls. Means must be provided on the... power unit. [Doc. No. 26344, 58 FR 18974, Apr. 9, 1993] ...

14 CFR 29.1142 - Auxiliary power unit controls.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Auxiliary power unit controls. 29.1142... AIRCRAFT AIRWORTHINESS STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Powerplant Controls and Accessories § 29.1142 Auxiliary power unit controls. Means must be provided on the flight deck for starting...

Control of feedwater composition of BWR power plant

International Nuclear Information System (INIS)

Sturla, P.; D'Anna, A.; Borgese, D.

1983-01-01

Corrosion behaviour of fuel element cladding, cycle structural materials and dose rate increase are relevant to physico-chemical characteristics of process coolants and to adopted operational conditions. A careful control of cycle chemistry, during loading and shutdown periods, is necessary to verify material choices, the polishing system and chemistry specifications. For this purpose ENEL carried out some preliminary experimental tests employing continuous control system and samples for specific analytical determinations. The cycle points checked during about two months were: main condensate; condensate after polishing system; outlet of low pressure heathers; final feedwater; inlet and outlet of clean-up system; drains to condenser. The physico-chemical analysis were related to corrosion product levels (Cu, Fe, Ni, Co) and water chemistry (pH, conductivity, dissolved oxygen etc.). The preliminary results allow to express some considerations about sampling procedures, detection limits and reliability of analytical employed methods. The acquisition data time and some morphological oxide pictures are also showed. (author)

Dedicated auxiliary power units for Hybrid Electric Vehicles

NARCIS (Netherlands)

Mourad, S.; Weijer, C.J.T. van de

1998-01-01

The use of a dedicated auxiliary power unit is essential to utilize the potential that hybrid vehicles offer for efficient and ultra-clean transportation. An example of a hybrid project at the TNO Road-Vehicles Research Institute shows the development and the results of a dedicated auxiliary power

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Nuclear reactors with auxiliary boiler circuit

International Nuclear Information System (INIS)

George, B.V.; Cook, R.K.

1976-01-01

A gas-cooled nuclear reactor has a main circulatory system for the gaseous coolant incorporating one or more main energy converting units, such as gas turbines, and an auxiliary circulatory system for the gaseous coolant incorporating at least one steam generating boiler arranged to be heated by the coolant after its passage through the reactor core to provide steam for driving an auxiliary steam turbine, such an arrangement providing a simplified start-up procedure also providing emergency duties associated with long term heat removal on reactor shut down

Transport in Auxiliary Heated NSTX Discharges

International Nuclear Information System (INIS)

LeBlanc, B.P.; Bell, M.G.; Bell, R.E.; Bitte, M.L.; Bourdelle, C.; Gates, D.A.; Kaye, S.M.; Maingi, R.; Menard, J.E.; Mueller, D.; Ono, M.; Paul, S.F.; Redi, M.H.; Roquemore, A.L.; Rosenberg, A.; Sabbagh, S.A.; Stutman, D.; Synakowski, E.J.; Soukhanovskii, V.A.; Wilson, J.R.

2003-01-01

The NSTX spherical torus (ST) provides a unique platform to investigate magnetic confinement in auxiliary-heated plasmas at low aspect ratio. Auxiliary power is routinely coupled to ohmically heated plasmas by deuterium neutral-beam injection (NBI) and by high-harmonic fast waves (HHFW) launch. While theory predicts both techniques to preferentially heat electrons, experiment reveals the electron temperature is greater than the ion temperature during HHFW, but the electron temperature is less than the ion temperature during NBI. In the following we present the experimental data and the results of transport analyses

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)

Trojan Nuclear Plant. 1976 Annual operating report

International Nuclear Information System (INIS)

1977-01-01

Net electrical energy generated was 2,102,986 MWH with the generator on line 3144.4 hours. Information is presented concerning operations, tests, corrective maintenance, auxiliary feedwater pump modifications, main generator fault, fuel performance, shutdowns and power outages, radiation exposure of personnel, radioactive and chemical effluents, and environmental surveillance

A CMOS pressure sensor tag chip for passive wireless applications.

Science.gov (United States)

Deng, Fangming; He, Yigang; Li, Bing; Zuo, Lei; Wu, Xiang; Fu, Zhihui

2015-03-23

This paper presents a novel monolithic pressure sensor tag for passive wireless applications. The proposed pressure sensor tag is based on an ultra-high frequency RFID system. The pressure sensor element is implemented in the 0.18 µm CMOS process and the membrane gap is formed by sacrificial layer release, resulting in a sensitivity of 1.2 fF/kPa within the range from 0 to 600 kPa. A three-stage rectifier adopts a chain of auxiliary floating rectifier cells to boost the gate voltage of the switching transistors, resulting in a power conversion efficiency of 53% at the low input power of -20 dBm. The capacitive sensor interface, using phase-locked loop archietcture, employs fully-digital blocks, which results in a 7.4 bits resolution and 0.8 µW power dissipation at 0.8 V supply voltage. The proposed passive wireless pressure sensor tag costs a total 3.2 µW power dissipation.

PS auxiliary magnet

CERN Multimedia

CERN PhotoLab

1974-01-01

Units of the PS auxiliary magnet system. The picture shows how the new dipoles, used for vertical and horizontal high-energy beam manipulation, are split for installation and removal so that it is not necessary to break the accelerator vacuum. On the right, adjacent to the sector valve and the windings of the main magnet, is an octupole of the set.

The English Primary Auxiliary Verbs: A Linguistic Theoretical Exercise

African Journals Online (AJOL)

Nekky Umera

Abstract. Obviously, the fact remains that English Language is a sensitive Language ... Even though the English auxiliary verbs are of two kinds: Primary and Modal auxiliary ..... Therefore we are of the opinion that most speakers lack adequate.

Detection of a regulating valve closure failure during review of recorded data after an automatic reactor shut down. Incident at the NPP Beznau-1, 27 April 1995

International Nuclear Information System (INIS)

Deutschmann, H.

1996-01-01

After recognizing a leak in the oil system of the running main feedwater pump 1 during rated power operation of the plant the operator changed feedwater supply manually to the stand-by pump 2. A short time later pump 2 was automatically tripped by the signal ''low oil pressure''. Immediate reduction of the reactor power by the operator was not successful because the scram signal ''low steam generator level and mismatch of steam/feedwater flow'' occurred and scram was actuated. In this plant a special operating feature, actuated by the scram signal, is implemented to reduce steam release to atmosphere in case of scram. The signal ''scram and average primary Temperature >287 deg. C opens the feedwater regulating valves, and later, if the average primary temperature decreases to <287 deg. C, they reclose by a redundant signal. In the experienced event, after the scram actuation, in the steam generator A a feedwater overfill occurred. The overfill protection tripped the operating feedwater pumps (main feedwater pump 3 and two auxiliary feedwater pumps). The large injection of water produced an overcooling of the primary with isolation of the volume control system outlet of the primary. The operator repaired the defective oil coolers of the feedwater pumps and restarted the plant. At that time, he had not recognized, that the plant response, which caused the steam generator overfill, was wrong. One day later, as all the recorded data were reviewed in more detail, it was found that the closure time of the feedwater regulating valve to steam generator A was much longer than designed (19 s instead 7 s). The operator requested an LCO for continued operation in spite of the fact, that the closure time was not fixed in the Technical specification. 3 figs

Operational experience on reduction of feedwater iron and liquid radwaste input for Kuosheng Nuclear Power Plant

International Nuclear Information System (INIS)

Wen, T.J.; Huang, Theresa Chen; Liu, Wen Tsung; Liu, T.C.; Shyur, Tzu Sheng; Shen, S.C.

1998-01-01

Other than cobalt alloys, or low cobalt materials, feedwater iron content plays an important role in crud activation and transport causing the growth of out-of-core radiation fields and associated with radwaste generation. Before installing prefilter in the upstream of condensate deep-bed demineralizer, increasing demand for suspended solid removal required new backwash and regeneration technique in Kuosheng Nuclear Power Plant. At steady state full power operation, the average iron concentration in condensate demineralizer influent was 8-15 ppb. Considering both the necessity of backwash and reduction of liquid radwaste input, several actions had been taken to promote the crud removal capabilities without using ultrasonic resin cleaner and controlled feedwater iron content between 0.5 and 2.0 ppb. This modified resin backwash technique would also generate minimum liquid radwaste. Meanwhile, significant efforts have been made to promote the quality of waste water by carefully control input streams as well as backwash modification to reduce liquid radwaste generation. The daily quantity of liquid radwaste has decreased dramatically in the past two years and is effectively controlled under the expected average daily input of design basis. (author)

Major factors in critical equipment reliability - Auxiliary systems; The development of an auxiliary system

International Nuclear Information System (INIS)

Forsthoffer, W.E.

1992-01-01

In this article, the author details the development of an actual auxiliary system in order to fully understand the function of each major component and how it contributes to the total operation and reliability of the system. Only after the function of an auxiliary system is thoroughly understood, can one proceed to discuss specifications, design audits, testing, operation and preventive maintenance. The application selected will be to develop a pressurized lubrication and steam turbine control oil system for the critical equipment unit. This example was selected since many readers will be familiar with this type and because it provides a good foundation towards understanding fluid sealing systems. In the exercise that follow, he will define the system requirements and determine the system parameters. This information will then be used for component sizing

Auxiliary bearing design considerations for gas cooled reactors

International Nuclear Information System (INIS)

Penfield, S.R. Jr.; Rodwell, E.

2001-01-01

The need to avoid contamination of the primary system, along with other perceived advantages, has led to the selection of electromagnetic bearings (EMBs) in most ongoing commercial-scale gas cooled reactor (GCR) designs. However, one implication of magnetic bearings is the requirement to provide backup support to mitigate the effects of failures or overload conditions. The demands on these auxiliary or 'catcher' bearings have been substantially escalated by the recent development of direct Brayton cycle GCR concepts. Conversely, there has been only limited directed research in the area of auxiliary bearings, particularly for vertically oriented turbomachines. This paper explores the current state-of-the-art for auxiliary bearings and the implications for current GCR designs. (author)

Energy consumption of auxiliary systems of electric cars

Directory of Open Access Journals (Sweden)

Evtimov Ivan

2017-01-01

Full Text Available The paper analyzes the power demand of the auxiliary systems of electric cars. On the basis of existing electric cars an analysis of energy consumption of different auxiliary systems is done. As a result possibilities for rational use of these systems have been proposed, which can increase the mileage per one charge of the battery.

Optimization of feed water control for auxiliary boiler

International Nuclear Information System (INIS)

Li Lingmao

2004-01-01

This paper described the feed water control system of the auxiliary boiler steam drum in Qinshan Phase III Nuclear Power Plant, analyzed the deficiency of the original configuration, and proposed the optimized configuration. The optimized feed water control system can ensure the stable and safe operation of the auxiliary boiler, and the normal operation of the users. (author)

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

2011-01-01

This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

Cleaning the feed-water pipeline internal surfaces

International Nuclear Information System (INIS)

Podkopaev, V.A.

1984-01-01

The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washing by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones

Cleaning the feed-water pipeline internal surfaces

Energy Technology Data Exchange (ETDEWEB)

Podkopaev, V.A.

1984-12-01

The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washed by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water with the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones.

Discussion of the concept of safety indicators from the point of view of TfUX2 accident sequence for Forsmark 3

International Nuclear Information System (INIS)

Bujor, A.

1991-01-01

This paper contains general considerations on the safety indicators, with details at the system level and for the operator actions. For the system analysis, a modular analysis at a low detailed level is proposed (Module System Approach) in order to emphasize the safety related aspects at the subsystem (module) level. The operator actions are divided in ''active actions'' (actions in the control room during incident/accident situations) and ''passive actions'' (actions during tests, maintenance, repairs, etc.) and are analysed separately. In the second part, a discussion of a possible way to apply some SI to the TfUX2 accident sequence for FORSMARK-3, is done. For the analysis of the Auxiliary Feedwater Systems (AFWS) an equation is proposed to derive target values for the failure probability on demand at the train level, given the target value at the system level, including the common cause failures between the redundant trains. (author) 6 tabs., 18 refs

Modernization of the feedwater heaters control level of the Almaraz I Nuclear Power Plant by OVATION system

International Nuclear Information System (INIS)

Madronal Rodriguez, E.; Cabrero Munoz, J. E.

2010-01-01

As a result of the process of technological renovation of the heaters system and the power increase project, Almaraz Nuclear Power Plant has made several design changes in the feedwater heaters system. Within these changes, the old heaters control loops are replaced because the new power will increase the heaters drainage caudal. This modernization is carried out using the OVATION control system.

Parallel Auxiliary Space AMG Solver for $H(div)$ Problems

Energy Technology Data Exchange (ETDEWEB)

Kolev, Tzanio V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Vassilevski, Panayot S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2012-12-18

We present a family of scalable preconditioners for matrices arising in the discretization of $H(div)$ problems using the lowest order Raviart--Thomas finite elements. Our approach belongs to the class of “auxiliary space''--based methods and requires only the finite element stiffness matrix plus some minimal additional discretization information about the topology and orientation of mesh entities. Also, we provide a detailed algebraic description of the theory, parallel implementation, and different variants of this parallel auxiliary space divergence solver (ADS) and discuss its relations to the Hiptmair--Xu (HX) auxiliary space decomposition of $H(div)$ [SIAM J. Numer. Anal., 45 (2007), pp. 2483--2509] and to the auxiliary space Maxwell solver AMS [J. Comput. Math., 27 (2009), pp. 604--623]. Finally, an extensive set of numerical experiments demonstrates the robustness and scalability of our implementation on large-scale $H(div)$ problems with large jumps in the material coefficients.

Auxiliary partial liver transplantation

NARCIS (Netherlands)

C.B. Reuvers (Cornelis Bastiaan)

1986-01-01

textabstractIn this thesis studies on auxiliary partial liver transplantation in the dog and the pig are reported. The motive to perform this study was the fact that patients with acute hepatic failure or end-stage chronic liver disease are often considered to form too great a risk for successful

Iron concentration controller in feedwater in nuclear plant

International Nuclear Information System (INIS)

Aizawa, Motohiro; Isaka, Yoshitaka

1990-01-01

The purpose of the present invention is to prevent chlorine ions from flowing into a reactor when sea water leakage accident should occur in a condenser upon control of Fe concentration in feedwater. That is, a sensor is disposed for detecting the leakage of the sea water at the exit of the condenser. The controller receives a detection signal as the input and delivers a control signal as the output. A control system receives the control signal and actuates valves in bypass systems. In view of the above, the electroconductivity or chlorine ion concentration of the condensate, which varies upon occurrence of sea water leakages in the condenser, is detected by the sensor, and then the controller closes a valve dispposed in the bypass systems in a processing device for filtering and desalting the condensates. Accordingly, the chlorine ions mixed into the condensates are removed by a desalting device without flowing into the reactor. In view of the above, an effect capable of keeping integrity of the plant is obtainable. (I.S.)

Design and scope of impact of auxiliary lanes : technical report.

Science.gov (United States)

2014-06-01

For decades, Texas Department of Transportation districts have constructed auxiliary lanes to support interchange : ramp operations and to resolve congestion proximate to freeway entrance and exit ramps. While auxiliary lanes are : built throughout T...

Boiler feedwater quality improvement by replacing conventional pre-treatment with advanced membrane systems

Energy Technology Data Exchange (ETDEWEB)

Doll, Bernhard [Process Systems Pall GmbH, Dreieich (Germany). Marketing; Venkatadri, Ramraj [Pall Corporation, Port Washington, NY (United States). Global Marketing Energy

2013-09-01

Two case studies in different application fields highlight significant economical and operational improvements that were achieved by replacing conventional water treatment technologies by highly-sophisticated membrane systems. The first case study deals with boiler feedwater in a power plant, focusing on the challenges faced as well as the direct and indirect benefits gained by the new system within a utility station. The second case study deals with the conventional water treatment scheme for groundwater from 13 wells at a major oil sands facility. Operational performance as well as the cost improvements gained in both cases will be presented. (orig.)

Probabilistic cloning with supplementary information contained in the quantum states of two auxiliary systems

International Nuclear Information System (INIS)

Li, Lvjun; Qiu, Daowen

2007-01-01

In probabilistic cloning with two auxiliary systems, we consider and compare three different protocols for the success probabilities of cloning. We show that, in certain circumstances, it may increase the success probability to add an auxiliary system to the probabilistic cloning machine having one auxiliary system, but we always can find another cloning machine with one auxiliary system having the same success probability as that with two auxiliary systems

TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

International Nuclear Information System (INIS)

Meier, J.K.

1983-01-01

The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

Device for detecting the water leak within a feedwater nozzle in water cooled reactors

International Nuclear Information System (INIS)

Hattori, Tsunekazu.

1984-01-01

Purpose: To enable exact recognition and detection for the state of water leak. Constitution: The detection device comprises a thermocouple disposed to the outer surface of a feedwater nozzle, a distortion meter for detecting the change in the outer diameter of a nozzle and an acoustic emission generator disposed to the inside of the nozzle for generating a signal upon temperature change. These sensors previously monitor the states during normal operation, and thus detect the change in each of the states upon occurrence of water leakage to issue an alarm. (Kamimura, M.)

Adaptation of computer code ALMOD 3.4 for safety analyses of Westighouse type NPPs and calculation of main feedwater loss

International Nuclear Information System (INIS)

Kordis, I.; Jerele, A.; Brajak, F.

1986-01-01

The paper presents theoretical foundations of ALMOD 3.4 code and modification done in order to adjust the model to westinghouse type NPP. test cases for verification of added modules functioning were done and loss of main feedwater (FW) transient at nominal power was analysed. (author)

Optimal Design of Complex Passive-Damping Systems for Vibration Control of Large Structures: An Energy-to-Peak Approach

Directory of Open Access Journals (Sweden)

Francisco Palacios-Quiñonero

2014-01-01

Full Text Available We present a new design strategy that makes it possible to synthesize decentralized output-feedback controllers by solving two successive optimization problems with linear matrix inequality (LMI constraints. In the initial LMI optimization problem, two auxiliary elements are computed: a standard state-feedback controller, which can be taken as a reference in the performance assessment, and a matrix that facilitates a proper definition of the main LMI optimization problem. Next, by solving the second optimization problem, the output-feedback controller is obtained. The proposed strategy extends recent results in static output-feedback control and can be applied to design complex passive-damping systems for vibrational control of large structures. More precisely, by taking advantages of the existing link between fully decentralized velocity-feedback controllers and passive linear dampers, advanced active feedback control strategies can be used to design complex passive-damping systems, which combine the simplicity and robustness of passive control systems with the efficiency of active feedback control. To demonstrate the effectiveness of the proposed approach, a passive-damping system for the seismic protection of a five-story building is designed with excellent results.

Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

International Nuclear Information System (INIS)

Hernandez M, J.L.; Ortiz V, J.

2005-01-01

The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

Eigenstates with the auxiliary field method

Energy Technology Data Exchange (ETDEWEB)

Semay, Claude [Service de Physique Nucleaire et Subnucleaire, Universite de Mons-UMONS, 20 Place du Parc, 7000 Mons (Belgium); Silvestre-Brac, Bernard, E-mail: claude.semay@umons.ac.b, E-mail: silvestre@lpsc.in2p3.f [LPSC Universite Joseph Fourier, Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, Avenue des Martyrs 53, F-38026 Grenoble-Cedex (France)

2010-07-02

The auxiliary field method is a powerful technique to obtain approximate closed-form energy formulas for eigenequations in quantum mechanics. Very good results can be obtained for Schroedinger and semirelativistic Hamiltonians with various potentials, even in the case of many-body problems. This method can also provide approximate eigenstates in terms of well-known wavefunctions, for instance harmonic oscillator or hydrogen-like states, but with a characteristic size which depends on quantum numbers. In this paper, we consider two-body Schroedinger equations with linear, logarithmic and exponential potentials and show that analytical approximations of the corresponding eigenstates can be obtained with the auxiliary field method, with very good accuracy in some cases.

Eigenstates with the auxiliary field method

International Nuclear Information System (INIS)

Semay, Claude; Silvestre-Brac, Bernard

2010-01-01

The auxiliary field method is a powerful technique to obtain approximate closed-form energy formulas for eigenequations in quantum mechanics. Very good results can be obtained for Schroedinger and semirelativistic Hamiltonians with various potentials, even in the case of many-body problems. This method can also provide approximate eigenstates in terms of well-known wavefunctions, for instance harmonic oscillator or hydrogen-like states, but with a characteristic size which depends on quantum numbers. In this paper, we consider two-body Schroedinger equations with linear, logarithmic and exponential potentials and show that analytical approximations of the corresponding eigenstates can be obtained with the auxiliary field method, with very good accuracy in some cases.

CRBRP decay heat removal systems

International Nuclear Information System (INIS)

Hottel, R.E.; Louison, R.; Boardman, C.E.; Kiley, M.J.

1977-01-01

The Decay Heat Removal Systems for the Clinch River Breeder Reactor Plant (CRBRP) are designed to adequately remove sensible and decay heat from the reactor following normal shutdown, operational occurrences, and postulated accidents on both a short term and a long term basis. The Decay Heat Removal Systems are composed of the Main Heat Transport System, the Main Condenser and Feedwater System, the Steam Generator Auxiliary Heat Removal System (SGAHRS), and the Direct Heat Removal Service (DHRS). The overall design of the CRBRP Decay Heat Removal Systems and the operation under normal and off-normal conditions is examined. The redundancies of the system design, such as the four decay heat removal paths, the emergency diesel power supplies, and the auxiliary feedwater pumps, and the diversities of the design such as forced circulation/natural circulation and AC Power/DC Power are presented. In addition to overall design and system capabilities, the detailed designs for the Protected Air Cooled Condensers (PACC) and the Air Blast Heat Exchangers (ABHX) are presented

S-matrix for the theories that admit closure of the algebra with the aid of auxiliary fields. Auxiliary fields in supergravity. [Word identities

Energy Technology Data Exchange (ETDEWEB)

Fradkin, E S; Vasiliev, M A [AN SSSR, Moscow. Fizicheskij Inst.

1978-08-19

A minimal set of auxiliary fields (scalarpseudoscalar and pseudovector) providing the closed algebra in supergravity is constructed. A compact scheme for the generating functional with closed gauge algebra is proposed. The S-matrix and the Ward identities for arbitrary theory that admits the closing of the algebra by introducing auxiliary fields is obtained.

Wigner Transport Simulation of Resonant Tunneling Diodes with Auxiliary Quantum Wells

Science.gov (United States)

Lee, Joon-Ho; Shin, Mincheol; Byun, Seok-Joo; Kim, Wangki

2018-03-01

Resonant-tunneling diodes (RTDs) with auxiliary quantum wells ( e.g., emitter prewell, subwell, and collector postwell) are studied using a Wigner transport equation (WTE) discretized by a thirdorder upwind differential scheme. A flat-band potential profile is used for the WTE simulation. Our calculations revealed functions of the auxiliary wells as follows: The prewell increases the current density ( J) and the peak voltage ( V p ) while decreasing the peak-to-valley current ratio (PVCR), and the postwell decreases J while increasing the PVCR. The subwell affects J and PVCR, but its main effect is to decrease V p . When multiple auxiliary wells are used, each auxiliary well contributes independently to the transport without producing side effects.

Operation auxiliary system (SAO)

International Nuclear Information System (INIS)

Lolich, J.; Santome, D.; Drexler, J.

1990-01-01

This work presents an auxiliary system for nuclear power plants operation (SAO). The development purpose consisted in a computing supervision system to be installed at different sites of a reactor, mainly in the control room. The inclusion of this system to a nuclear power plant minimizes the possibility of human error for the facility operation. (Author) [es

Passive houses in Sweden. Experiences from design and construction phase

Energy Technology Data Exchange (ETDEWEB)

Jansson, Ulla

2008-07-01

The sector of residential buildings and service organizations uses 36% of the total energy in Sweden. In June 2006, it was decided by the Swedish parliament that the energy use in residential buildings and premises should decrease by 20% per heated unit area before 2020. To reach this goal, more energy efficient buildings must be produced as well as energy efficient improvements must be performed on the existing building stock. One way to reduce the energy use in buildings is to build passive houses. A passive house is a mechanically ventilated building that with a highly insulated and air tight building envelope uses a minimum of energy for heating. The method used in this research is to practically participate in four passive house demonstration projects. The results expected are to find guiding principles and tools needed for passive house planning and make the system solutions usable for planning in more general terms. Joining as a part of the planning group; advice and help is given to architects, consultants and to the client. The demonstration projects studied are located in the south-west of Sweden. Three of the projects are new constructions and one is a renovation project. In the centre of Vanirom 40 rental apartments were built in 2005/2006 according to the passive house standard. Solar collectors on the roof contribute to the domestic hot water production. Every apartment has its own mechanical ventilation system with efficient heat recovery. Auxiliary heating is supplied by electricity. The load bearing structure were made of concrete and cast at site. The exterior walls and roof were made of wooden frame construction and mounted at site. The tenants moved in during summer 2006. The passive house project in Failure's consists of three houses with 12 rental apartments. The air is supplied by mechanical ventilation with an air to air heat exchanger, one in each apartment. The domestic hot water is prepared by solar collectors and auxiliary heating

Auxiliary bearing design and rotor dynamics analysis of blower fan for HTR-10

International Nuclear Information System (INIS)

Gao Mingshan; Yang Guojun; Xu Yang; Zhao Lei; Yu Suyuan

2005-01-01

The electromagnetic bearing instead of ordinary mechanical bearing was chosen to support the rotor in the blower fan system with helium of 10 MW high temperature gas-cooled test reactor (HTR-10), and the auxiliary bearing was applied in the HTR-10 as the backup protector. When the electromagnetic bearing doesn't work suddenly for the power broken, the auxiliary bearing is used to support the falling rotor with high rotating speed. The rotor system will be protected by the auxiliary bearing. The design of auxiliary bearing is the ultimate safeguard for the system. This rotor is vertically mounted to hold the blower fan. The rotor's length is about 1.5 m, its weight is about 240 kg and the rotating speed is about 5400 r/min. Auxiliary bearing design and rotor dynamics analysis are very important for the design of blower fan to make success. The research status of the auxiliary bearing was summarized in the paper. A sort of auxiliary bearing scheme was proposed. MSC.Marc was selected to analyze the vibration mode and the natural frequency of the rotor. The scheme design of auxiliary bearing and analysis result of rotor dynamics offer the important theoretical base for the protector design and control system of electromagnetic bearing of the blower fan. (authors)

The Range of Gapping and the Status of Auxiliaries.

Science.gov (United States)

Warner, A. R.

Full verbs and auxiliaries are subject to gapping. In the simplest cases, this construction type involves apparent ellipsis within one or more clausal conjuncts under identity with the finite verb or auxiliary of a preceding conjunct. It has often been suggested that the apparent ellipsis must involve at least a verb. Some researchers see in the…

Auxiliary office chair

OpenAIRE

Pascual Osés, Maite

2007-01-01

The aim of this project is to develop an auxiliary office chair, which favorably will compete with the existing chairs on the market. Evolutions of ergonomical survey in the work environment and on the configuration of offices require new products which fulfill the requirements properly. In order to achieve it a survey about office chairs has been carried out: types, characteristics, ways of usage and products on the market besides a large antropometrical study and ergonomics related to work ...

The relationship among the solutions of two auxiliary ordinary differential equations

International Nuclear Information System (INIS)

Liu Xiaoping; Liu Chunping

2009-01-01

In a recent article [Phys. Lett. A 356 (2006) 124], Sirendaoreji extended their auxiliary equation method by introducing a new auxiliary ordinary differential equation (NAODE) and its 14 solutions. Then the author studied some nonlinear evolution equations (NLEEs) and got more exact travelling wave solutions. In this paper, we will show that the 14 solutions of the NAODE are actually the same as the solutions obtained by original auxiliary equation method, and they are only different in the form.

Simplified technique for auxiliary orthotopic liver transplantation using a whole graft

Science.gov (United States)

ROCHA-SANTOS, Vinicius; NACIF, Lucas Souto; PINHEIRO, Rafael Soares; DUCATTI, Liliana; ANDRAUS, Wellington; D'ALBURQUERQUE, Luiz Carneiro

2015-01-01

Background Acute liver failure is associated with a high mortality rate and the main purposes of treatment are to prevent cerebral edema and infections, which often are responsible for patient death. The orthotopic liver transplantation is the gold standard treatment and improves the 1-year survival. Aim To describe an alternative technique to auxiliary liver transplant on acute liver failure. Method Was performed whole auxiliary liver transplantation as an alternative technique for a partial auxiliary liver transplantation using a whole liver graft from a child removing the native right liver performed a right hepatectomy. The patient met the O´Grady´s criteria and the rational to indicate an auxiliary orthotopic liver transplantation was the acute classification without hemodynamic instability or renal failure in a patient with deterioration in consciousness. Results The procedure improved liver function and decreased intracranial hypertension in the postoperative period. Conclusion This technique can overcome some postoperative complications that are associated with partial grafts. As far as is known, this is the first case of auxiliary orthotopic liver transplantation in Brazil. PMID:26176253

Window-mounted auxiliary solar heater

Science.gov (United States)

Anthony, K. G.; Herndon, E. P.

1977-01-01

System uses hot-air collectors, no thermal storage, and fan with thermostat switches. At cost of heating efficiency, unit could be manufactured and sold at price allowing immediate entry to market as auxiliary heating system. Its simplicity allows homeowner installation, and maintenance is minimal.

Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

International Nuclear Information System (INIS)

Wang, T.-C.; Wang, S.-J.; Teng, J.-T.

2005-01-01

This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

Auxiliary basis expansions for large-scale electronic structure calculations.

Science.gov (United States)

Jung, Yousung; Sodt, Alex; Gill, Peter M W; Head-Gordon, Martin

2005-05-10

One way to reduce the computational cost of electronic structure calculations is to use auxiliary basis expansions to approximate four-center integrals in terms of two- and three-center integrals, usually by using the variationally optimum Coulomb metric to determine the expansion coefficients. However, the long-range decay behavior of the auxiliary basis expansion coefficients has not been characterized. We find that this decay can be surprisingly slow. Numerical experiments on linear alkanes and a toy model both show that the decay can be as slow as 1/r in the distance between the auxiliary function and the fitted charge distribution. The Coulomb metric fitting equations also involve divergent matrix elements for extended systems treated with periodic boundary conditions. An attenuated Coulomb metric that is short-range can eliminate these oddities without substantially degrading calculated relative energies. The sparsity of the fit coefficients is assessed on simple hydrocarbon molecules and shows quite early onset of linear growth in the number of significant coefficients with system size using the attenuated Coulomb metric. Hence it is possible to design linear scaling auxiliary basis methods without additional approximations to treat large systems.

Analogue to digital upgrade project-boiler feedwater control system for Bruce Power nuclear units 1 & 2

International Nuclear Information System (INIS)

Long, R.

2012-01-01

Bruce Power Nuclear Generating Station A, “Bruce A” is in the final stages of its Restart Project. This capital project will see a large scale rehabilitation of Units 1 and 2 resulting in addition of 1500MW of safe, reliable, clean electricity to the Ontario grid. Restart Project Scope 375, Boiler Feedwater Controls Upgrade was sanctioned to replace obsolete analog devices with a modern digital control system. This project replaced the existing Foxboro H Line analog controls which comprised of 81 individual control modules and support instrumentation. The replacement system was a Triconex Triple Modular Redundant PLC which interfaces with two redundant touch screen monitors. The upgraded digital system incorporates the following controls: 1. Boiler Level Control Loops 2. Dearator Level Control Loops 3. Dearator Pressure Control Loops 4. Boiler Feedwater Recirculation Flow Control Loops A number of technical challenges were addressed when installing a new digital system within the existing plant configuration. Interfaces to new, old and refurbished field devices must be understood as well as implications of connecting to the plant’s Digital Control Computers (DCC’s) and newly installed Steam Generators. The overall project involved many stakeholders to address various requirements from conceptual / design stage through procurement, construction, commissioning and return to service. In addition, the project highlighted the unique requirements found in Nuclear Industry with respect to Human Factors and Software Quality Assurance. (author)

ISP 22 OECD/NEA/CSNI International standard problem n. 22. Evaluation of post-test analyses

International Nuclear Information System (INIS)

1992-07-01

The present report deals with the open re-evaluation of the originally double-blind CSNI International Standard Problem 22 based on the test SP-FW-02 performed in the SPES facility. The SPES apparatus is an experimental simulator of the Westinghouse PWR-PUN plant. The test SP-FW-02 (ISP22) simulates a complete loss of feedwater with delayed injection of auxiliary feedwater. The main parts of the report are: outline of the test facility and of the SP-FW-02 experiment; overview of pre-test activities; overview of input models used by post-test participants; evaluation of participant predictions; evaluation of qualitative and quantitative code accuracy of pre-test and post-test calculations

Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

International Nuclear Information System (INIS)

Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

1993-01-01

The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

Safety device and machine system of nuclear power plant

International Nuclear Information System (INIS)

1978-10-01

It introduces principle and kinds of heat power including heat balance and nuclear power. It explains a lot of technical terms about the nuclear power system, which are primary loop, reactor, steam generator, primary coolant pump and pressurizer in PWR, chemical and volume control system, component cooling system, safety injection system, and spent fuel cooling and storage system in auxiliary system, liquid solid and gaseous waste disposal system in radwaste disposal, gland sealing system, turbine instrumentation, turning gear, hydrogen cooling system, condenser, feedwater heater, degenerate heater, auxiliary heat exchanger, centrifugal pump, rotary reciprocating and tank and pressure vessel.

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

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

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

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

Study on decay heat removal capability of reactor vessel auxiliary cooling system

International Nuclear Information System (INIS)

Nishi, Y.; Kinoshita, I.

1991-01-01

The reactor vessel auxiliary cooling system (RVACS) is a simple, Passive decay heat removal system for an LMFBR. However, the heat removal capacity of this system is small compared to that of an immersed type of decay heat exchanger. In this study, a high-porosity porous body is proposed to enhance the RVACS's heat transfer performance to improve its applicability. The objectives of this study are to propose a new method which is able to use thermal radiation effectively, to confirm its heat removal capability and to estimate its applicability limit of RVACS for an LMFBR. Heat transfer tests were conducted in an experimental facility with a 3.5 m heat transfer height to evaluate the heat transfer performance of the high-porosity porous body. Using the experimental results, plant transient analyses were performed for a 300 MWe pool type LMFBR under a Total Black Out (TBO) condition to confirm the heat removal capability. Furthermore, the relationship between heat removal capability and thermal output of a reactor were evaluated using a simple parameter model

Goal programming for cyclical auxiliary police scheduling at UiTM Cawangan Perlis

Science.gov (United States)

Mustapar, Wasilatul Effah; Nasir, Diana Sirmayunie Mohd; Nor, Nor Azriani Mohamad; Abas, Sharifah Fhahriyah Syed

2017-11-01

Constructing a good and fair schedule for shift workers poses a great challenge as it requires a lot of time and effort. In this study, goal programming has been applied on scheduling to achieve the hard and soft constraints for a cyclical schedule that would ease the head of auxiliary police at building new schedules. To accomplish this goal, shift types were assigned in order to provide a fair schedule that takes into account the auxiliary police's policies and preferences. The model was run using Lingo software. Three out of four goals set for the study were achieved. In addition, the results considered an equal allocation for every auxiliary police, namely 70% for total duty and 30% for the day. Furthermore, the schedule was able to cyclically generate another 10 sets schedule. More importantly, the model has provided unbiased scheduling of auxiliary policies which led to high satisfaction in auxiliary police management.

Minimal set of auxiliary fields and S-matrix for extended supergravity

Energy Technology Data Exchange (ETDEWEB)

Fradkin, E S; Vasiliev, M A [Physical Lebedev Institute - Moscow

1979-05-19

Minimal set of auxiliary fields for linearized SO(2) supergravity and one-parameter extension of the minimal auxiliary fields in the SO(1) supergravity are constructed. The expression for the S-matrix in SO(2) supergravity are given.

BE, DO, and Modal Auxiliaries of 3-Year-Old African American English Speakers

Science.gov (United States)

Newkirk-Turner, Brandi L.; Oetting, Janna B.; Stockman, Ida J.

2014-01-01

Purpose: This study examined African American English--speaking children's use of BE, DO, and modal auxiliaries. Method: The data were based on language samples obtained from 48 three-year-olds. Analyses examined rates of marking by auxiliary type, auxiliary surface form, succeeding element, and syntactic construction and by a number of child…

Leak Detection of High Pressure Feedwater Heater Using Empirical Models

International Nuclear Information System (INIS)

Lee, Song Kyu; Kim, Eun Kee; Heo, Gyun Young; An, Sang Ha

2009-01-01

Even small leak from tube side or pass partition within the high pressure feedwater heater (HPFWH) causes a significant deficiency in its performance. Plant operation under the HPFWH leak condition for long time will result in cost increase. Tube side leak within HPFWH can produce the high velocity jet of water and it can cause neighboring tube failures. However, most of plants are being operated without any information for internal leaks of HPFWH, even though it is prone to be damaged under high temperature and high pressure operating conditions. Leaks from tubes and/or pass partition of HPFWH occurred in many nuclear power plants, for example, Mihama PS-2, Takahama PS-2 and Point Beach Nuclear Plant Unit 1. If the internal leaks of HPFWH are monitored, the cost can be reduced by inexpensive repairs relative to loss in performance and moreover plant shutdown as well as further tube damages can be prevented

Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2

International Nuclear Information System (INIS)

Tijerina S, F.

2008-01-01

The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

Modified Darboux transformations with foreign auxiliary equations

International Nuclear Information System (INIS)

Schulze-Halberg, Axel

2011-01-01

We construct a new type of first-order Darboux transformations for the stationary Schroedinger equation. In contrast to the conventional case, our Darboux transformations support arbitrary (foreign) auxiliary equations. We show that among other applications, our formalism can be used to systematically construct Darboux transformations for Schroedinger equations with energy-dependent potentials, including a recent result (Lin et al., 2007) as a special case. -- Highlights: → We generalize the Darboux transformation for the Schroedinger equation. → By admitting arbitrary auxiliary functions, we provide a new tool for generating solutions. → As a special case we recover a recent result on energy-dependent potentials. → We extend the latter result to very general energy-dependence.

Nuclear plant power up-rate study: feedwater heater evaluations

International Nuclear Information System (INIS)

Svensson, Eric; Catapano, Michael; Coakley, Michael; Thomas, Dan

2014-01-01

Given today's nuclear industry business climate, it has become common for Utility companies to consider increasing unit capacities through turbine replacement and power up-rates. An integral part of the studies conducted by many towards this end, involve the generation of a set of turbine cycle heat balances with predicted performance parameters for the up-rated condition. Once these tentative operating values are established, it becomes necessary to evaluate the suitability of the existing components within each system to ensure they are capable of continued safe and reliable operation. The ultimate cost for the up-rate, including the cost for any major required modifications or significant replacements is weighed against increased revenue generated from the up-rate over time. Exelon's Peach Bottom Atomic Power Station (PBAPS) is currently planning for an Extended Power up-rate (EPU) for both units. To ensure the existing Feedwater Heaters (FWH) could maintain the operating and transient response margins at the EPU condition, an engineering study was conducted. Powerfect Inc. in conjunction with SPX Heat Transfer LLC were contracted to provide engineering services to analyze the design, thermal performance, reliability and operating conditions at projected EPU conditions. Specifically, to address the following with regard to the station's Feedwater Heaters (FWHs): 1. Evaluate Drain Cooler (DC) Velocities - including zone inlet velocity, cross and window velocities and outlet velocities. 2. Evaluate Drain Cooler Zone Pressure Drop for effect on drain cooler margins to flashing. 3. Evaluate differential pressure allowable across the pass partition plate. 4. Evaluate Drain Cooler Tube Vibration Potential. 5. Perform detailed steam dome velocity calculations. The goal of the study was to identify the most susceptible areas within the heaters for problems and potential failures when operating at the higher duty of the EPU condition for the remaining life

Dummy auxiliaries in child and adult second language acquisition of Dutch

NARCIS (Netherlands)

Blom, W.B.T.; de Korte, S.

2011-01-01

In previous research it has been observed that second language (L2) learners of Dutch and German use analytic verbs in contexts where the target language requires synthetic verbs. These analytic verbs consist of a semantically empty auxiliary (dummy auxiliary) that selects a lexical infinitive. In

Extensions of the auxiliary field method to solve Schroedinger equations

International Nuclear Information System (INIS)

Silvestre-Brac, Bernard; Semay, Claude; Buisseret, Fabien

2008-01-01

It has recently been shown that the auxiliary field method is an interesting tool to compute approximate analytical solutions of the Schroedinger equation. This technique can generate the spectrum associated with an arbitrary potential V(r) starting from the analytically known spectrum of a particular potential P(r). In the present work, general important properties of the auxiliary field method are proved, such as scaling laws and independence of the results on the choice of P(r). The method is extended in order to find accurate analytical energy formulae for radial potentials of the form aP(r) + V(r), and several explicit examples are studied. Connections existing between the perturbation theory and the auxiliary field method are also discussed

Extensions of the auxiliary field method to solve Schroedinger equations

Energy Technology Data Exchange (ETDEWEB)

Silvestre-Brac, Bernard [LPSC Universite Joseph Fourier, Grenoble 1, CNRS/IN2P3, Institut Polytechnique de Grenoble, Avenue des Martyrs 53, F-38026 Grenoble-Cedex (France); Semay, Claude; Buisseret, Fabien [Groupe de Physique Nucleaire Theorique, Universite de Mons-Hainaut, Academie universitaire Wallonie-Bruxelles, Place du Parc 20, B-7000 Mons (Belgium)], E-mail: silvestre@lpsc.in2p3.fr, E-mail: claude.semay@umh.ac.be, E-mail: fabien.buisseret@umh.ac.be

2008-10-24

It has recently been shown that the auxiliary field method is an interesting tool to compute approximate analytical solutions of the Schroedinger equation. This technique can generate the spectrum associated with an arbitrary potential V(r) starting from the analytically known spectrum of a particular potential P(r). In the present work, general important properties of the auxiliary field method are proved, such as scaling laws and independence of the results on the choice of P(r). The method is extended in order to find accurate analytical energy formulae for radial potentials of the form aP(r) + V(r), and several explicit examples are studied. Connections existing between the perturbation theory and the auxiliary field method are also discussed.

Experimental fast reactor JOYO MK-III functional test. Primary auxiliary cooling system test

International Nuclear Information System (INIS)

Karube, Koji; Akagi, Shinji; Terano, Toshihiro; Onuki, Osamu; Ito, Hideaki; Aoki, Hiroshi; Odo, Toshihiro

2004-03-01

This paper describes the results of primary auxiliary cooling system, which were done as a part of JOYO MK-III function test. The aim of the tests was to confirm the operational performance of primary auxiliary EMP and the protection system including siphon breaker of primary auxiliary cooling system. The items of the tests were: (Test No.): (Test item). 1) SKS-117: EMP start up test. 2) SKS-118-1: EMP start up test when pony motor running. 3) SKS-121: Function test of siphon breaker. The results of the tests satisfied the required performance, and demonstrated successful operation of primary auxiliary cooling system. (author)

Development of Intelligent Auxiliary System for Customized Physical Fitness and Healthcare

Directory of Open Access Journals (Sweden)

Huang Chung-Chi

2016-01-01

Full Text Available With the advent of global high-tech industry and commerce era, the sedentary reduces opportunities of physical activity. And physical fitness and health of people is getting worse and worse. At present, the shortage of physical fitness instructors greatly affected the effectiveness of health promotion. Therefore, it is necessary to develop an auxiliary system which can reduce the workload of instructors and enhance physical fitness and health for people. But current general physical fitness and healthcare system is hard to meet individualized needs. The main purpose of this research is to develop an intelligent auxiliary system for customized physical fitness and healthcare. It records all processes of physical fitness and healthcare system by wireless sensors network. The results of intelligent auxiliary systems for customized physical fitness and healthcare will be generated by fuzzy logic Inference. It will improve individualized physical fitness and healthcare. Finally, we will demonstrate the advantages of the intelligent auxiliary system for customized physical fitness and healthcare.

Passive Strategy with Integrated Passive Safety System (IPSS) for DBAs in SBO

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho; Kim, Jihee; Choi, Jae Young; Jeon, Inseop; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

Beznau nuclear power plant. The third solution is to make a cooling redundancy. Integrated passive safety system (IPSS) proposed in 2011 is a representative design adding big water tanks outside containment on a top of auxiliary building as a heat sink and water supplier. Three methods are each merit and demerit. After the design concept of IPSS was proposed, main functions were analyzed to be verified. With the simulated performances of IPSS in conservative conditions, passive counter strategies to cope with design based accidents (DBAs) were proposed and described in a basic condition of a SBO.

Aspectual auxiliary verbs in Xitsonga

African Journals Online (AJOL)

Kate H

Let him always come' e. á vá hátl-é vá yá étlélà. OPT 3PL quickly-OPT 3PL go sleep. 'Let them quickly go to bed'. 3.4 Negative markers. Negation is marked on AA verbs. The auxiliary verb hatla 'quickly' is negated in three tenses.

A Polyethylene Moderator Design for Auxiliary Ex-core Neutron Detector

International Nuclear Information System (INIS)

Lee, Hwan Soo; Shin, Ho Cheol; Bae, Seong Man

2012-01-01

The moderator of detector assembly in ENFMS (Excore Neutron Flux Monitoring System) plays a key role for slowing down from fast neutron to thermal neutron at outside of reactor vessel. Since neutron monitoring detector such as BF3, fission chamber detectors mostly responds to thermal neutron, moderator should be included to neutron detector assembly to detect more efficiently. Generally, resin has been used for moderator of detector in ENFMS of OPR1000 and APR1400, because resin has stable thermal resistance, availability and high neutron moderation characteristics due to the light atomic materials. In case of an auxiliary ex-core neutron detector, the polyethylene is suggested that polyethylene has a better moderator rather than resin, then, the amounts of moderator are reduced. This is important thing for auxiliary ex-core detector equipment at reactor, because the auxiliary equipment should affect minimally to another system. In this study, polyethylene moderator is designed for auxiliary ex-core neutron detector. To find out the optimal thickness of polyethylene moderator, preliminary simulation and experiments are performed. And sensitivity simulation for detector moderator at actual reactor is performed by DORT code

An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

International Nuclear Information System (INIS)

Di Salvo, R.; Leonard, M.T.; Wreathall, J.

1986-01-01

An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

An initial examination of aging related degradation in turbine drives and governors for safety related pumps

International Nuclear Information System (INIS)

Cox, D.F.

1991-01-01

This study is being performed to examine the relationship between time dependent degradation, and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized water reactor (PWR) plants, and the Reactor Core Isolation Cooling (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in NPRDS, reviewing Licensee Event Reports, and thoroughly investigating contacts with operating plant personnel, and by personal observation. The reported information was reviewed to determine the cause of the event and the method of discovery. From this data attempts have been made at determining the predictability of events and possible preventive measures that may be implemented. Findings in a recent study on the Auxiliary Feedwater System (NUREG/CR-5404) indicate that the turbine drive is the single largest contributor to AFW system degradation. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design seem to indicate that this equipment can be a reliable component in safety systems

Theoretical and numerical analysis of auxiliary heating for cryogenic target fabrication

International Nuclear Information System (INIS)

Yang Xiaohu; Tian Chenglin; Yin Yan; Xu Han; Zhuo Hongbin

2008-01-01

In order to compensate for the nonspherical-symmetric heat flux in the hohlraum, auxiliary heating is usually applied to the outside wall of the hohlraum during the cooling process. A one-dimensional heat exchange theoretical model has been proposed in the indirect-drive target, to analyze the required auxiliary heat flux. With a two dimensional axisymmetric model, the auxiliary heating mechanism has been simulated by FLUENT code. The optimum heat flux which is 635 W/m 2 has been obtained as the heaters around the outside of the hohlraum about 1.3 mm above and below the mid-plane. The result is in good agreement with the theoretical model. (authors)

Entanglement Capacity of Two-Qubit Unitary Operator with the Help of Auxiliary System

International Nuclear Information System (INIS)

Hu Baolin; Di Yaomin

2007-01-01

The entanglement capacity of general two-qubit unitary operators is studied when auxiliary systems are allowed, and the analytical results based on linear entropy when input states are disentangled are given. From the results the condition for perfect entangler, α 1 = α 2 = π/4, is obtained. Contrary to the case without auxiliary system, the parameter α 3 may play active role to the entanglement capacity when auxiliary systems are allowed.

The Future Mission Tasking and Resourcing of the U.S. Coast Guard Auxiliary

Science.gov (United States)

2012-09-01

the years. Therefore, this study cannot present what the demographic makeup of the Auxiliary was at any past point, and cannot identify if there were...Auxiliary, and the Nation’s projected demographic composition. The demographic makeup of the country, the country’s future volunteers, is changing...command in the spring of 2010.104 This one-page document broad- brushes three prioritized overarching Auxiliary missions: 1) to promote and improve

Environmental practices of the auxiliary companies to the Spanish automobile industry

Science.gov (United States)

González-Torre, Pilar L.; González, Beatriz A.; Gupta, Surendra M.

2005-11-01

The automobile manufacturing industry plays a very important role in a country's economy. The importance of automobile manufacturing industry lies in its sheer size and complexity in terms of the direct and indirect influence it commands across many other industries. While millions of people are employed in the automobile manufacturing industry, it is estimated that more than two and half times that number are employed in the auxiliary companies that supply parts to the automobile manufacturing companies. The auxiliary companies represent a group of businesses of various sizes, types, and geographical locations, producing a vast variety of products ranging from the very simple to the extremely intricate. In this study, the current environmental practices of management in the core Spanish auxiliary companies that do business with the automobile manufacturing industry (and thus form a large part of the automobile manufacturing industry's supply chain) are investigated. We show that while automobile manufacturing companies are under scrutiny to become more and more environmentally friendly, not only at their manufacturing stage but also at their products' useful and EOL stages, there appears to be no such burden on the auxiliary companies. Our conclusion is based on an elaborate survey conducted during the fall of 2004 of Spanish auxiliary companies with questions about the characteristics, environmental practices and reverse logistics related activities carried out by the companies.

Curricular Guidelines for Dental Auxiliary Radiology.

Science.gov (United States)

Journal of Dental Education, 1981

1981-01-01

AADS curricular guidelines suggest objectives for these areas of dental auxiliary radiology: physical principles of X-radiation in dentistry, related radiobiological concepts, principles of radiologic health, radiographic technique, x-ray films and intensifying screens, factors contributing to film quality, darkroom, and normal variations in…

Linearized curvatures for auxiliary fields in the de Sitter space

Energy Technology Data Exchange (ETDEWEB)

Vasiliev, M A

1988-09-19

New consistent linearized curvatures in the de Sitter space are constructed. The sequence of actions, describing bosonic and fermionic gauge auxiliary fields, is found based on these curvatures. The proposed actions are parametrized by two integer parameters, n greater than or equal to 0 and m greater than or equal to 0. The simplest case n=m=0 corresponds in the flat limit to the auxiliary fields of 'new minimal' supergravity. The hamiltonian formulation is developed for the auxiliary fields suggested; hamiltonians and first- and second-class constraints are constructed. Using these results, it is shown that the systems of fields proposed possess no dynamical degrees of freedom in de Sitter and flat spaces. In addition the hamiltonian formalism is analysed for some free dynamical systems based on linearized higher-spin curvatures introduced previously.

Improvement of stability of sinusoidally driven atmospheric pressure plasma jet using auxiliary bias voltage

Directory of Open Access Journals (Sweden)

Hyun-Jin Kim

2015-12-01

Full Text Available In this study, we have proposed the auxiliary bias pulse scheme to improve the stability of atmospheric pressure plasma jets driven by an AC sinusoidal waveform excitation source. The stability of discharges can be significantly improved by the compensation of irregular variation in memory voltage due to the effect of auxiliary bias pulse. From the parametric study, such as the width, voltage, and onset time of auxiliary bias pulse, it has been demonstrated that the auxiliary bias pulse plays a significant role in suppressing the irregular discharges caused by the irregular variation in memory voltage and stable discharge can be initiated with the termination of the auxiliary bias pulse. As a result of further investigating the effects of the auxiliary pulse scheme on the jet stability under various process conditions such as the distance between the jet head and the counter electrode, and carrier gas flow, the jet stability can be improved by adjusting the amplitude and number of the bias pulse depending on the variations in the process conditions.

Auxiliary/Master microprocessor CAMAC Crate Controller applications

International Nuclear Information System (INIS)

Barsotti, E.

1975-01-01

The need for further sophistication of an already complex serial CAMAC control system at Fermilab led to the development of an Auxilary/Master CAMAC Crate Controller. The controller contains a Motorola 6800 microprocessor, 2K bytes of RAM, and 8K bytes of PROM memory. Bussed dataway lines are time shared with CAMAC signals to provide memory expansion and direct addressing of peripheral devices without the need of external cabling. The Auxiliary/Master Crate Controller (A/MCC) can function as either a Master, i.e., stand alone, crate controller or as an Auxiliary controller to Fermilab's Serial Crate Controller (SCC). Two modules, one single- and one double-width, make up an A/MCC. The microprocessor has one nonmaskable and one maskable vectored interrupt. Time sharing the dataway between SCC programmed and block transfer generated dataway cycles and A/MCC operations still allows a 99 percent microprocessor CPU busy time. Since the conception of the A/MCC, there has been an increasing number of control system-related projects proposed which would not have been possible or would have been very difficult to implement without such a device. The first such application now in use at Fermilab is a stand-alone control system for a mass spectrometer experiment in the Main Ring Internal Target Area. This application in addition to other proposed A/MCC applications, both stand-alone and auxiliary, is discussed

Improving Semi-Supervised Learning with Auxiliary Deep Generative Models

DEFF Research Database (Denmark)

Maaløe, Lars; Sønderby, Casper Kaae; Sønderby, Søren Kaae

Deep generative models based upon continuous variational distributions parameterized by deep networks give state-of-the-art performance. In this paper we propose a framework for extending the latent representation with extra auxiliary variables in order to make the variational distribution more...... expressive for semi-supervised learning. By utilizing the stochasticity of the auxiliary variable we demonstrate how to train discriminative classifiers resulting in state-of-the-art performance within semi-supervised learning exemplified by an 0.96% error on MNIST using 100 labeled data points. Furthermore...

Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

International Nuclear Information System (INIS)

1980-01-01

The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

Modal Auxiliaries and Their Semantic Functions Used by Advanced EFL Learners

Science.gov (United States)

Torabiardakani, Najmeh; Khojasteh, Laleh; Shokrpour, Nasrin

2015-01-01

Since modal auxiliary verbs have been proved to be one of the most troublesome grammatical structures in English, the researchers of this study decided to do an analysis on the ways in which advanced EFL Iranian students use modal auxiliaries focusing specially on nine modals' semantic functions. Consequently, was conducted based on the following…

Assessment of System Behavior and Actions Under Loss of Electric Power For CANDU

Energy Technology Data Exchange (ETDEWEB)

Kang, San Ha; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

2014-05-15

For the analysis, the CANDU-6 plant in Korea is considered and only the passive components are operable. The other systems are assumed to be at failed condition due to the loss of electric power. At this accident, only the inventories remained in the primary heat transport system (PHTS) and steam generator can be used for the decay heat removal. Due to the transfer of decay heat, the inventory of steam generator secondary side is discharged to the air through passive operation of main steam safety valves (MSSVs). After the steam generators are dried, the PHTS is over-pressurized and the coolant is discharged to fuelling machine vault through passive operation of degasser condenser tank relief valves (DCRVs). Under this situation, the maintenance of the integrity of PHTS is important for the protection of radionuclides release to the environment. Thus, deterministic analysis using CATHENA code is carried out for the simulation of the accident and the appropriate operator action is considered. The loss of electric power results in the depletion of steam generator inventory which is necessary for the decay heat removal. If only the passive system is credited, the PT can be failed after the steam generator is depleted. For the prevention of the PT failure, the feedwater should be supplied to the steam generator before 4,800s after the accident. The feedwater can be supplied using water in dousing tank if the steam generators are depressurized. The decay heat from the core is removed through natural circulation if the feedwater can be supplied continuously.

Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

International Nuclear Information System (INIS)

Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

2008-01-01

The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

Analysis of limit cycling on a boiler feedwater control system

International Nuclear Information System (INIS)

Thomas, P.J.; Harrison, T.A.; Hollywell, P.D.

1986-01-01

During operation of the UKAEA Prototype Fast Reactor, it was found that oscillations sometimes occurred in the boiler feedwater systems. These were normally of relatively low amplitude, but led to the adoption of low controller gains so that control was rather slack. While control performance proved generally adequate for steady running, the lack of tight control of steam drum levels sometimes led to difficulties during periods when plant conditions were undergoing major change. The paper discusses the methods used to gain a full understanding of the phenomena occurring, and describes how that knowledge is being used to improve the control system so as to eliminate the limit cycling modes and ensure good control of steam drum levels. A noteworthy feature of the study was the use of two independent representations of plant behaviour: (i) a frequency response model, FWRFREQ, and (ii) a time-domain simulation model, PFRTDM. The simplified analysis of FWRFREQ proved to be of enormous value in identifying modes of system behaviour; PFRTDM was used as a detailed check on the accuracy and validity of the results obtained. (author)

Development of KALIMER auxiliary sodium and cover gas management system

International Nuclear Information System (INIS)

Kwon, Sang Woon; Hwang, Sung Tae

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author)

Development of KALIMER auxiliary sodium and cover gas management system

Energy Technology Data Exchange (ETDEWEB)

Kwon, Sang Woon; Hwang, Sung Tae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-11-01

The objectives of this report are to develop and to describe the auxiliary liquid metal and cover gas management systems of KALIMER. the system includes following system: (1) Auxiliary liquid metal system (2) Inert gas receiving and processing system (3) Impurity monitoring and analysis system. Auxiliary liquid metal and cover gas management system of KALIMER was developed. Functions of each systems and design basis were describes. The auxiliary liquid metal system receives, transfers, and purifies all sodium used in the plant. The system furnishes the required sodium quantity at the pressure, temperature, flow rate, and purity specified by the interfacing system. The intermediated sodium processing subsystem (ISPS) provides continuous purification of IHTS sodium, as well as performs the initial fill operation for both the IHTS and reactor vessel. The primary sodium processing subsystem provides purification (cold trapping) for sodium used in the reactor vessel. The inert gas receiving and processing (IGRP) system provides liquefied and ambient gas storage, delivers inert gases of specified composition and purity at regulated flow rates and pressures to points of usage throughout the KALIMER, and accepts the contaminated gases through its vacuum facilities for storage and transfer to the gas radwaste system. Three gases are used in the KALIMER: helium, argon, and nitrogen. 11 tabs., 12 figs. (Author).

Hydraulic turbines and auxiliary equipment

Energy Technology Data Exchange (ETDEWEB)

Luo Gaorong [Organization of the United Nations, Beijing (China). International Centre of Small Hydroelectric Power Plants

1995-07-01

This document presents a general overview on hydraulic turbines and auxiliary equipment, emphasizing the turbine classification, in accordance with the different types of turbines, standard turbine series in China, turbine selection based on the basic data required for the preliminary design, general hill model curves, chart of turbine series and the arrangement of application for hydraulic turbines, hydraulic turbine testing, and speed regulating device.

Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

International Nuclear Information System (INIS)

Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

1984-01-01

This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

International Nuclear Information System (INIS)

Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Severe transient tests on operation steam generators: Analysis of the fluid structure dynamic thermal interaction

International Nuclear Information System (INIS)

Billon, F.; David, J.; Procaccia, H.

1983-01-01

The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculations of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant. This code divides the S.G. into three parts: - the first concerns the S.G.s region above the downcomer, zone where the mixing between hot water and cold feedwater occurs, - the second is the downcomer itself which is divided into n segments, - the third concerns the tube sheet zone which is also divided into n segments. The most severe transient test performed is the auxiliary cold feedwater injection into the steam generator during a hot standby of the plant: two levels of flow rate have been realised: 55 and 110 m 3 /h of 42 0 C feedwater. The tests have shown that if the cold feedwater injection occurs when the steam generator water level is below feedwater ring, the lowest fluid temperature reached at tube sheet inlet is about 230 0 C. (orig.)

Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

International Nuclear Information System (INIS)

Goldberg, A.; Streit, R.D.

1981-05-01

Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

Progress on radio frequency auxiliary heating system designs in ITER

International Nuclear Information System (INIS)

Makowski, M.; Bosia, G.; Elio, F.

1996-09-01

ITER will require over 100 MW of auxiliary power for heating, on- and off-axis current drive, accessing the H-mode, and plasma shut-down. The Electron Cyclotron Range of Frequencies (ECRF) and Ion Cyclotron Range of Frequencies (ICRF) are two forms of Radio Frequency (RF) auxiliary power being developed for these applications. Design concepts for both the ECRF and ICRF systems are presented, key features and critical design issues are discussed, and projected performances outlined

An auxiliary differential equation FDTD method for anisotropic magnetized plasmas

International Nuclear Information System (INIS)

Liu Shaobin; Mo Jinjun; Yuan Naichang

2004-01-01

An auxiliary differential equation finite-difference time-domain (ADE-FDTD) methodology for anisotropic magnetized plasmas is derived. The method is based on a difference approximation of the auxiliary differential equation. A comparison with the JEC method is included. The CPU time saving by several times and accuracy of the method are confirmed by computing the reflection and transmission through a magnetized plasma layer with the direction of propagation parallel to the direction of the biasing field

Electrospinning of aligned fibers with adjustable orientation using auxiliary electrodes

International Nuclear Information System (INIS)

Arras, Matthias M L; Grasl, Christian; Schima, Heinrich; Bergmeister, Helga

2012-01-01

A conventional electrospinning setup was upgraded by two turnable plate-like auxiliary high-voltage electrodes that allowed aligned fiber deposition in adjustable directions. Fiber morphology was analyzed by scanning electron microscopy and attenuated total reflection Fourier transform infrared spectroscopy (FTIR-ATR). The auxiliary electric field constrained the jet bending instability and the fiber deposition became controllable. At target speeds of 0.9 m s −1 90% of the fibers had aligned within 2°, whereas the angular spread was 70° without the use of auxiliary electrodes. It was even possible to orient fibers perpendicular to the rotational direction of the target. The fiber diameter became smaller and its distribution narrower, while according to the FTIR-ATR measurement the molecular orientation of the polymer was unaltered. This study comprehensively documents the feasibility of directed fiber deposition and offers an easy upgrade to existing electrospinning setups. (paper)

Aiming of Kirkpatrick-Baez microscope based on auxiliary optical system

International Nuclear Information System (INIS)

Huang Shengling; Mu Baozhong; Yi Shengzhen; Wang Xin; Wang Zhanshan; Ding Yongkun; Miao Wenyong; Dong Jianjun

2009-01-01

An auxiliary optical system has been designed, which can provide precise positioning for aiming Kirkpatrick-Baez (KB) microscope object location. An 8 keV X-ray imaging system by KB microscope with periodic multilayer films has been designed. The field of view and depth of field in the resolution of 5 μm are got, and then the corresponding point and depth of field in diagnostic experiments are calculated. Based on the object-image relations and precision of the KB microscope, an auxiliary visible light imaging system is designed and X-ray imaging experiments are performed, which can achieve equivalent aiming between the visible imaging system and the KB microscope. The results show that ±20 μm vertical axis plane and ±300 μm axial accuracy are achieved through the auxiliary optical path, which can meet the object point positioning requirements of the KB microscope. (authors)

14 CFR 33.96 - Engine tests in auxiliary power unit (APU) mode.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Engine tests in auxiliary power unit (APU... TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: AIRCRAFT ENGINES Block Tests; Turbine Aircraft Engines § 33.96 Engine tests in auxiliary power unit (APU) mode. If the engine is designed with a propeller brake which...

30 CFR 57.22208 - Auxiliary fans (I-A, II-A, III, and V-A mines).

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Auxiliary fans (I-A, II-A, III, and V-A mines... fans (I-A, II-A, III, and V-A mines). (a) Auxiliary fans, except fans used in shops and other areas... applicable requirements of 30 CFR part 18, and be operated so that recirculation is minimized. Auxiliary fans...

Transient Performance of Air-cooled Condensing Heat Exchanger in Long-term Passive Cooling System during Decay Heat Load

Energy Technology Data Exchange (ETDEWEB)

Kim, Myoung Jun; Lee, Hee Joon [Kookmin University, Seoul (Korea, Republic of); Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In [KAERI, Daejeon (Korea, Republic of)

2015-05-15

In the event of a 'loss of coolant accident'(LOCA) and a non-LOCA, the secondary passive cooling system would be activated to cool the steam in a condensing heat exchanger that is immersed in an emergency cooldown tank (ECT). Currently, the capacities of these ECTs are designed to be sufficient to remove the sensible and residual heat from the reactor coolant system for 72 hours after the occurrence of an accident. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. Therefore, the tank should be refilled regularly from an auxiliary water supply system when the system is used for more than 72 hours. Otherwise, the system would fail to dissipate heat from the condensing heat exchanger due to the loss of the cooling water. Ultimately, the functionality of the passive cooling system would be seriously compromised. As a passive means of overcoming the water depletion in the tank, Kim et al. applied for a Korean patent covering the concept of a long-term passive cooling system for an ECT even after 72 hours. This study presents transient performance of ECT with installing air-cooled condensing heat exchanger under decay heat load. The cooling capacity of an air-cooled condensing heat exchanger was evaluated to determine its practicality.

Bayesian Analysis of Geostatistical Models With an Auxiliary Lattice

KAUST Repository

Park, Jincheol

2012-04-01

The Gaussian geostatistical model has been widely used for modeling spatial data. However, this model suffers from a severe difficulty in computation: it requires users to invert a large covariance matrix. This is infeasible when the number of observations is large. In this article, we propose an auxiliary lattice-based approach for tackling this difficulty. By introducing an auxiliary lattice to the space of observations and defining a Gaussian Markov random field on the auxiliary lattice, our model completely avoids the requirement of matrix inversion. It is remarkable that the computational complexity of our method is only O(n), where n is the number of observations. Hence, our method can be applied to very large datasets with reasonable computational (CPU) times. The numerical results indicate that our model can approximate Gaussian random fields very well in terms of predictions, even for those with long correlation lengths. For real data examples, our model can generally outperform conventional Gaussian random field models in both prediction errors and CPU times. Supplemental materials for the article are available online. © 2012 American Statistical Association, Institute of Mathematical Statistics, and Interface Foundation of North America.

46 CFR 182.620 - Auxiliary means of steering.

Science.gov (United States)

2010-10-01

... TONS) MACHINERY INSTALLATION Steering Systems § 182.620 Auxiliary means of steering. (a) Except as... personnel hazards during normal or heavy weather operation. (b) A suitable hand tiller may be acceptable as...

Auxiliary Armed Forces and Innovations in Security Governance in Mozambique’s Civil War

NARCIS (Netherlands)

Jentzsch, C.

2017-01-01

Who rules during the civil war? This article argues that the concept of armed group governance must be expanded to include auxiliary armed forces linked to rebels or the government. Comparing the organization of rebel and government auxiliaries, the article demonstrates that security governance

Stereoselective conjugate radical additions: application of a fluorous oxazolidinone chiral auxiliary for efficient tin removal.

Science.gov (United States)

Hein, Jason E; Zimmerman, Jake; Sibi, Mukund P; Hultin, Philip G

2005-06-23

[reaction: see text] A series of asymmetric free-radical-mediated intermolecular conjugate additions using a fluorous oxazolidinone chiral auxiliary has been completed. The fluorous auxiliary facilitated product isolation using fluorous solid phase extractions (FSPE), effectively removing excess organic and organometallic reagents. Parallel reactions carried out with a similar but nonfluorous norephedrine-derived oxazolidinone demonstrated the superior stereoselectivity and purification obtainable with the fluorous chiral auxiliary.

Auxiliary power unit for moving a vehicle

Science.gov (United States)

Akasam, Sivaprasad [Peoria, IL; Johnson, Kris W [Peoria, IL; Johnson, Matthew D [Peoria, IL; Slone, Larry M [Washington, IL; Welter, James Milton [Chillicothe, IL

2009-02-03

A power system is provided having at least one traction device and a primary power source configured to power the at least one traction device. In addition, the power system includes an auxiliary power source also configured to power the at least one traction device.

An Auxiliary Equation for the Bellman Equation in a One-Dimensional Ergodic Control

International Nuclear Information System (INIS)

Fujita, Y.

2001-01-01

In this paper we consider the Bellman equation in a one-dimensional ergodic control. Our aim is to show the existence and the uniqueness of its solution under general assumptions. For this purpose we introduce an auxiliary equation whose solution gives the invariant measure of the diffusion corresponding to an optimal control. Using this solution, we construct a solution to the Bellman equation. Our method of using this auxiliary equation has two advantages in the one-dimensional case. First, we can solve the Bellman equation under general assumptions. Second, this auxiliary equation gives an optimal Markov control explicitly in many examples

TMI-2 auxiliary building elevator shaft and pit decontamination

Energy Technology Data Exchange (ETDEWEB)

Bengel, T.G.

1986-01-01

Decontamination of the elevator pit and shaft in the auxiliary building at Three Mile Island Unit 2 (TMI-2) was performed to remove high radiation and contamination levels which prevented personnel from utilizing the elevator. The radiation and contamination levels in the TMI-2 auxiliary building elevator shaft have been reduced to the point where plant personnel are again permitted to ride in the elevator without a radiation work permit, with the exception of access to the 281-ft (basement) level. Based on the declassification and expanded use of the elevator, the task goal has been met. The tax expended 16.16 man-rem and 621 man-hours.

Treatment for chronic daily headache by using auxiliary and alternative methods

Directory of Open Access Journals (Sweden)

V. A. Golovacheva

2015-01-01

Full Text Available Chronic daily headache (CDH is one of the top 10 causes of adult disability and one of the 5 most common causes of female disability. To treat patients with CDH is one of the most difficult tasks in neurological practice. Difficulties in managing patients with CHD are associated with the high prevalence of comorbid mental disorders, analgesic abuse, pain syndromes at another site, and misconceptions of a patient about his/her disease. A combination of drug and non-drug therapies is the mainstay of the current approach to treating patients with CDH. Standard, alternative, and auxiliary therapies are identified. The paper describes different types of current auxiliary and alternative therapy used in the world’s leading headache centers and clinics. It describes experience with cerebrolysin used as auxiliary and alternative pharmacotherapies for CDH.

Design and Experiment of Auxiliary Bearing for Helium Blower of HTR-PM

International Nuclear Information System (INIS)

Yang Guojun; Shi Zhengang; Liu Xingnan; Zhao Jingjing

2014-01-01

The helium blower is the important equipment for HTR-PM. Active magnetic bearing (AMB) instead of mechanical bearing is selected to support the rotor of the helium blower. However, one implication of AMB is the requirement to provide the auxiliary bearing to mitigate the effects of failures or overload conditions. The auxiliary bearing is used to support the rotor when the AMB fails to work. It must support the dropping rotor and bear the great impact force and friction heat. The design of the auxiliary bearing is one of the challenging problems in the whole system. It is very important for the helium blower with AMB of HTR-PM to make success. The rotor’s length of helium blower of HTR-PM is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The axial load is 4500kg, and the radial load is 1950kg. The angular contact ball bearing was selected as the auxiliary bearing. The test rig has been finished. It is difficult to analyze the falling course of the rotor. The preliminary analysis of the dropping rotor was done in the special condition. The impact force of auxiliary bearing was computed for the axial and radial load. And the dropping test of the blower rotor for HTR-10 will be introduced also in this paper. Results offer the important theoretical base for the protector design of the helium blower with AMB for HTR-PM. (author)

New set of auxiliary fields for supergravity theories

International Nuclear Information System (INIS)

Oliveira Rivelles, V. de.

1983-02-01

A brief introduction on supersymmetry is given. The problems with the obtainment of the auxiliary fields in supergravity theories are discussed, after a short presentation of the supersymmetry algebra representations. (L.C.) [pt

76 FR 28795 - Privacy Act of 1974; Department of Homeland Security United States Coast Guard-024 Auxiliary...

Science.gov (United States)

2011-05-18

... 1974; Department of Homeland Security United States Coast Guard-024 Auxiliary Database System of... Security/United States Coast Guard-024 Auxiliary Database (AUXDATA) System of Records.'' This system of...: United States Coast Guard Auxiliary Database (AUXDATA). Security classification: Unclassified. System...

Direction of Impurity Pinch and Auxiliary Heating in Tokamak Plasmas

International Nuclear Information System (INIS)

Angioni, C.; Peeters, A.G.

2006-01-01

A mechanism of particle pinch for trace impurities in tokamak plasmas, arising from the effect of parallel velocity fluctuations in the presence of a turbulent electrostatic potential, is identified analytically by means of a reduced fluid model and verified numerically with a gyrokinetic code for the first time. The direction of such a pinch reverses as a function of the direction of rotation of the turbulence in agreement with the impurity pinch reversal observed in some experiments when moving from dominant auxiliary ion heating to dominant auxiliary electron heating

Research needs for risk-informed, performance-based regulation

International Nuclear Information System (INIS)

Bailey, J.A.

1997-01-01

Palo Verde Nuclear Generating Station has used PRA-derived risk insights for about 10 years now. The plant originally started applying PRA modeling to an auxiliary feedwater system during the initial licensing phases of the plant, and as a result of that, they were able to work with the NRC and apply some graded quality requirements to that particular system. There was a third redundant auxiliary feedwater pump, and they now can treat that system as partially safety related and partially non-safety related. So it was an advance for Palo Verde at that time to be able to make decisions with a PRA and they began learning how to use those techniques. After completing the IPE it became natural for the plant to make a transition into other areas at the plant to look for areas where the insights gained from PRA could be applied into their decision-making processes. Those that the plant embarked upon initially were areas where they could gain operational risk assessment insights. The author goes on to discuss experiences gained in using these techniques to better assess the safety of operations within the plant. In addition he offers comments on areas which need further development and research to make them more applicable to a plant by plant basis

GEOTHERMAL / SOLAR HYBRID DESIGNS: USE OF GEOTHERMAL ENERGY FOR CSP FEEDWATER HEATING

Energy Technology Data Exchange (ETDEWEB)

Craig Turchi; Guangdong Zhu; Michael Wagner; Tom Williams; Dan Wendt

2014-10-01

This paper examines a hybrid geothermal / solar thermal plant design that uses geothermal energy to provide feedwater heating in a conventional steam-Rankine power cycle deployed by a concentrating solar power (CSP) plant. The geothermal energy represents slightly over 10% of the total thermal input to the hybrid plant. The geothermal energy allows power output from the hybrid plant to increase by about 8% relative to a stand-alone CSP plant with the same solar-thermal input. Geothermal energy is converted to electricity at an efficiency of 1.7 to 2.5 times greater than would occur in a stand-alone, binary-cycle geothermal plant using the same geothermal resource. While the design exhibits a clear advantage during hybrid plant operation, the annual advantage of the hybrid versus two stand-alone power plants depends on the total annual operating hours of the hybrid plant. The annual results in this draft paper are preliminary, and further results are expected prior to submission of a final paper.

Fatigue analysis of a PWR steam generator tube sheet

International Nuclear Information System (INIS)

Billon, F.; Buchalet, C.; Poudroux, G.

1985-01-01

The fatigue analysis of a PWR steam generator (S.G) tube sheet is threefold. First, the flow, pressure and temperature variations during the design transients are defined for both the primary fluid and the normal and auxiliary feedwater. Second, the flow, velocities, pressure and temperature variations of the secondary fluid at the bottom of the downcomer and above the tube sheet are determined for the transients considered. Finally, the corresponding temperatures and stresses in the tube sheet are calculated and the usage factors determined at various locations in the tube sheet. The currently available standard design transients for the primary fluid and the feedwater are too conservative to be utilized as such in the fatigue analysis of the S.G. tube sheets. Thus, a detailed examination and reappraisal of each operating transient was performed. The revised design conditions are used as inputs to the calculation model TEMPTRON. TEMPTRON determines the mixing conditions between the feedwater and the recirculation fluid from the S.G. feedwater nozzles to the center of the tube sheet via the downcomer. The fluid parameters, flow rate and velocity, temperature and pressure variations, as a function of the time during the transients are obtained. Finally, the usage factors at various locations on the tube sheet are derived using the standard ASME section III method

Renormalization of supersymmetric models without using auxiliary fields

International Nuclear Information System (INIS)

Urbanek, P.

1986-01-01

Previously a linear representation of supersymmetry (Ss) was used in investigations of renormalizability. There auxiliary fields have been introduced in order that the Ss-algebra closes 'off-shell'. When the auxiliary fields are eliminated by their equations of motion, the Ss representation becomes nonlinear and Ss closes only 'on-shell'. Following O.Piguet and K.Sibold 1984 Ss is expressed through Ward identities which are formulated as functional variations of the generating functional of the Green functions. These functional operators form a closed algebra, a fact essential for the proof of renormalizability, which is given. It is not necessary to use a specific subtraction scheme in the Green functions. The procedure is applied to the Wess-Zumino model and the supersymmetric extension of the quantum electrodynamics. 15 refs. (qui)

Auxiliary equipment cooling circuit in nuclear reactors

International Nuclear Information System (INIS)

Yanagisawa, Ko.

1986-01-01

Purpose: To prevent the propagation of bacterias that transform NO 2 into NO 3 in auxiliary equipment coolants using corrosion inhibitors of nitrite type in BWR type reactors. Method: In auxiliary equipments coolant systems, water quality is controlled by using purified water as supplement water and nitrite such as Na 2 NO 2 as the corrosion inhibitors. However, in the circumstance where dissolved oxygen is present, bacteria propagate to oxidize NO 2 into NO 3 . Thus, NO 2 at 200 ppm is reduced to 20 ppm. In view of the above, a surge tank supplied from water supplement line is connected in series and a deaeration device is disposed thereto. Since the presence of dissolved oxygen causes the bacteria to propagate it is desired that the dissolved oxygen density in the supplement water is less than 5 ppm. Deaeration and pressure reduction in the surge tank can remove the dissolved oxygen, prevent NO 3 increase and also prevent stress corrosion cracks in the system pipeways. (Horiuchi, T.)

LQG/LTR [linear quadratic Gaussian with loop transfer recovery] robust control system design for a low-pressure feedwater heater train

International Nuclear Information System (INIS)

Murphy, G.V.; Bailey, J.M.

1990-01-01

This paper uses the linear quadratic Gaussian with loop transfer recovery (LQG/LTR) control system design method to obtain a level control system for a low-pressure feedwater heater train. The control system performance and stability robustness are evaluated for a given set of system design specifications. The tools for analysis are the return ratio, return difference, and inverse return difference singular-valve plots for a loop break at the plant output. 3 refs., 7 figs., 2 tabs

Loss Model and Efficiency Analysis of Tram Auxiliary Converter Based on a SiC Device

Directory of Open Access Journals (Sweden)

Hao Liu

2017-12-01

Full Text Available Currently, the auxiliary converter in the auxiliary power supply system of a modern tram adopts Si IGBT as its switching device and with the 1700 V/225 A SiC MOSFET module commercially available from Cree, an auxiliary converter using all SiC devices is now possible. A SiC auxiliary converter prototype is developed during this study. The author(s derive the loss calculation formula of the SiC auxiliary converter according to the system topology and principle and each part loss in this system can be calculated based on the device datasheet. Then, the static and dynamic characteristics of the SiC MOSFET module used in the system are tested, which aids in fully understanding the performance of the SiC devices and provides data support for the establishment of the PLECS loss simulation model. Additionally, according to the actual circuit parameters, the PLECS loss simulation model is set up. This simulation model can simulate the actual operating conditions of the auxiliary converter system and calculate the loss of each switching device. Finally, the loss of the SiC auxiliary converter prototype is measured and through comparison it is found that the loss calculation theory and PLECS loss simulation model is valuable. Furthermore, the thermal images of the system can prove the conclusion about loss distribution to some extent. Moreover, these two methods have the advantages of less variables and fast calculation for high power applications. The loss models may aid in optimizing the switching frequency and improving the efficiency of the system.

An Analytical Method of Auxiliary Sources Solution for Plane Wave Scattering by Impedance Cylinders

DEFF Research Database (Denmark)

Larsen, Niels Vesterdal; Breinbjerg, Olav

2004-01-01

Analytical Method of Auxiliary Sources solutions for plane wave scattering by circular impedance cylinders are derived by transformation of the exact eigenfunction series solutions employing the Hankel function wave transformation. The analytical Method of Auxiliary Sources solution thus obtained...

EP 1000 -The European Passive Plant

International Nuclear Information System (INIS)

Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

1998-01-01

A group of European utilities, along with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT) initiated a program to evaluate Westinghouse passive nuclear plant technology for application in Europe. The European utility group consisted of: Agrupacion electrica para al Desarrollo Technologico Nuclear (DTN), Spain; Electricite de France; ENEL, SpA., Italy; IVO Power Engineering, Ltd., Finland; Scottish Nuclear Limited (acting for itself on behalf of Nuclear Electric plc, U.K.; Tractebel Energy Engineering, Belgium; UAK (represented by NOK-Beznau), Switzerland; and Vattenfall AB, Ringhals, Sweden. The European Passive Plant (EPP) program, which began in 1994, is an evaluation of the Westinghouse 600 MWe AP 600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs in meeting the European Utility Requirements (EUR), and where necessary, modifying the design to achieve compliance. Phase 1 or the EPP program was completed and included the two major tasks of evaluating the effect of the EUR on the Westinghouse nuclear island and developing the EP 1000, a 1000 MWe passive plant reference design that conforms to the EUR and would be licensable in Europe. The EP 1000 closely follows the Westinghouse SPWR design for safety systems and containment and the AP 600 design for auxiliary systems. It also includes features that where required to meet the EUR and key European licensing requirements. The primary circuit of the EP 1000 retains most of the general features of the current-day designs, but some evolutionary features to enhance reliability, simplicity of operation, ease of maintenance, and plant safety have been incorporated into the design. The core, reactor vessel, and reactor internals of the EP 1000 are similar to those of currently operating Westinghouse PWR plants, but several new features are included to enhance the performance characteristics. The basic EP 1000 safety philosophy is based on use of inherent

CADRIGS--computer aided design reliability interactive graphics system

International Nuclear Information System (INIS)

Kwik, R.J.; Polizzi, L.M.; Sticco, S.; Gerrard, P.B.; Yeater, M.L.; Hockenbury, R.W.; Phillips, M.A.

1982-01-01

An integrated reliability analysis program combining graphic representation of fault trees, automated data base loadings and reference, and automated construction of reliability code input files was developed. The functional specifications for CADRIGS, the computer aided design reliability interactive graphics system, are presented. Previously developed fault tree segments used in auxiliary feedwater system safety analysis were constructed on CADRIGS and, when combined, yielded results identical to those resulting from manual input to the same reliability codes

The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

International Nuclear Information System (INIS)

Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

1992-07-01

This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

The sensitivity of auxiliary examinations in different stages of sporadic Creutzfeldt-Jakob disease

Directory of Open Access Journals (Sweden)

Jiao-jiao JIANG

2017-06-01

Full Text Available Objective To analyze the sensitivity of auxiliary examinations in different periods of sporadic Creutzfeldt-Jakob disease (sCJD. Methods The clinical data of 53 sCJD patients were retrospectively analyzed including the different stages of skull diffusion-weighted magnetic resonance imaging (DWI, 24-hour ambulatory electroencephalogram (EEG, 18F-FDG PET/CT (PET-CT and cerebrospinal fluid 14-3-3 protein. When calculating the sensitivity of an auxiliary examination, the diagnostic criteria were defined by combining the specific clinical manifestations with two or more positive results of other auxiliary examinations. Results There were 24, 53 and 22 sCJD patients, respectively, met the criterion of early (E, middle (M and later (L stage of disease (some patients fit 2 or 3 stages. The sensitivity of DWI (E: 58.3%, M: 85.4%, L: 94.7%, EEG (E: 45.8%, M: 62.7%, L: 77.8%, 14-3-3 protein in cerebrospinal fluid (E: 11.1%, M: 52.9% and PET-CT (E: 80%, M: 100% increased gradually with disease progression. The sensitivity of PET-CT was higher than the other auxiliary examinations for E and M stages; no PET-CT was conducted in L stage. High signal regions mainly distributed in the cortex in E and M stages, but in L stage, no significant difference was found on the distribution of high signal regions between cortex and basal ganglia. Conclusions The sensitivities of the auxiliary examinations were different for sCJD patients in different stages. Reexaminations in different periods may improve the sensitivity for sCJD diagnosis. The sensitivity of PET-CT was high, and the combination of PET-CT and other auxiliary examinations may play a key role in the diagnosis of sCJD. DOI: 10.11855/j.issn.0577-7402.2017.05.15

Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

Energy Technology Data Exchange (ETDEWEB)

Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

2015-05-15

Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

Considerations on the question of applying ion exchange or reverse osmosis methods in boiler feedwater processing

International Nuclear Information System (INIS)

Marquardt, K.; Dengler, H.

1976-01-01

This consideration is to show that the method of reverse osmosis presents in many cases an interesting and economical alternative to part and total desolination plants using ion exchangers. The essential advantages of the reverse osmosis are a higher degree of automization, no additional salting of the removed waste water, small constructional volume of the plant as well as favourable operational costs with increasing salt content of the crude water to be processed. As there is a relatively high salt breakthrough compared to the ion exchange method, the future tendency in boiler feedwater processing will be more towards a combination of methods of reverse osmosis and post-purification through continuous ion exchange methods. (orig./LH) [de

ERP-IV-A program for transient thermal-hydraulic analysis of PWR plant

International Nuclear Information System (INIS)

Dai Anguo; Tang Jiahuan; Qian Huifu; Gao Zhikang

1987-12-01

The author deal with the descriptions of physical model of transient process in PWR plant and the function of ERP-IV (ERR-IV Transient Thermo-Hydraulic Analysis Code). The code has been developed for safety analysis and design transient. The code is characterized by the multi-loop long-term, short term, wide-range plant simulation with the capability to analyze natural circulation condition. The description of ERP-IV includes following parts: reactor, primary coolant loops, pressurizer, steam generators, main steam system, turbine, feedwater system, steam dump, relive valves, and safety valves in secondary side, etc.. The code can use for accident analysis, such as loss of all A.C. power to power plant auxiliaries (a station blackout), loss of normal feedwater, loss of load, loss of condenser vacuum and other events causing a turbine trip, complete loss of forced reactor coolant flow, uncontrolled rod cluster control assembly bank withdrawal. It can also be used for accident analysis of the emergency and limiting conditions, such as feedwater line break and main steam line rupture. It can also be utilized as a tool for system design studies, component design, setpoint studies and design transition studies, etc

Quick look report for semiscale MOD-2C Test S-FS-11

International Nuclear Information System (INIS)

Plessinger, M.P.

1985-11-01

Results of a preliminary analysis of the fifth test performed in the Semiscale MOD-2C Steam Generator Feedwater and Steam Line Break (FS) experiment series are presented. Test S-FS-11 simulated a pressurized water reactor transient initiated by a 50% break in a steam generator bottom feedwater line downstream of the check valve. With the exception of primary pressure, the initial conditions represented the initial conditions used for the C-E System 80 Final Safety Analysis Report (FSAR) Appendix 15B calculations. The transient included an initial 600 s period in which only automatic plant protection systems responded to the initiating event. This period was followed by a series of operator actions necessary to stabilize the plant followed by break isolation and affected loop steam generator refill with auxiliary feedwater. The test results provided a measured evaluation of the effectiveness of the automatic responses in minimizing primary system overpressurization and operator actions in stabilizing the plant. Test data also provided a basis for comparison with other tests in the series of the effects of break size on primary overpressurization and primary-to-secondary heat transfer. 64 figs

Fundamental studies of passivity and passivity breakdown

International Nuclear Information System (INIS)

Macdonald, D.D.; Urquidi-Macdonald, M.; Song, H.; Biaggio-Rocha, S.; Searson, P.

1991-11-01

This report summarizes the findings of our fundamental research program on passivity and passivity breakdown. During the past three and one half years in this program (including the three year incrementally-funded grant prior to the present grant), we developed and experimentally tested various physical models for the growth and breakdown of passive films on metal surfaces. These models belong to a general class termed ''point defects models'' (PDMs), in which the growth and breakdown of passive films are described in terms of the movement of anion and cation vacancies

Awareness of biomedical waste management among dental professionals and auxiliary staff in Amritsar, India.

Science.gov (United States)

Narang, Ramandeep S; Manchanda, Adesh; Singh, Simarpreet; Verma, Nitin; Padda, Sarfaraz

2012-12-01

The aim of this study was to determine awareness of biomedical waste (BMW) management policies and practices among dental professionals and auxiliary staff in a dental hospital/clinics in Amritsar, India, to inform the development of future policies for effective implementation of BMW rules. The study involved 160 staff members at the Amritsar hospital/clinics (80 dentists and 80 auxiliary staff) to whom a questionnaire was distributed regarding policies, practices and awareness relating to BMW. The questionnaire was first piloted. Completed questionnaires were returned anonymously. The resulting data were statistically tested using the chi-square test for differences between the dentists and auxiliary staff. In respect of BMW management policies, there was a highly significant difference in the responses of the dentists, whose answers suggested far greater knowledge than that of the auxiliaries (Pmanagement practices, the dentists were significantly more aware (Pwaste collection in the hospital and the disposal of various items into different colour-coded bags. As for employee education/awareness, there was a significant difference (Pmanagement among dental auxiliary staff in the dental hospital/clinics in Amritsar and a lack of awareness of some aspects among dentists who work in the hospital/clinics. The results provide the hospital authorities with data upon which they can develop a strategy for improving BMW management.

Auxiliary facilities on nuclear ship 'MUTSU'

International Nuclear Information System (INIS)

Tsujimura, Shotaro; Takigami, Yoshio.

1989-01-01

The nuclear ship 'MUTSU' has been moored at SEKINEHAMA, MUTU City in AOMORI Prefecture and several tests and works are being carried out on the ship. The construction of the auxiliary facilities for these works on the ship was completed in safety in August 1988. After that the facilities have fulfilled their function. The outlines of design, fabrication and construction of the facilities are described in this paper. (author)

Categorical Data Fusion Using Auxiliary Information

OpenAIRE

Fosdick, Bailey K.; DeYoreo, Maria; Reiter, Jerome P.

2015-01-01

In data fusion, analysts seek to combine information from two databases comprised of disjoint sets of individuals, in which some variables appear in both databases and other variables appear in only one database. Most data fusion techniques rely on variants of conditional independence assumptions. When inappropriate, these assumptions can result in unreliable inferences. We propose a data fusion technique that allows analysts to easily incorporate auxiliary information on the dependence struc...

Progress in the integration of the ITER plant systems in auxiliary buildings

International Nuclear Information System (INIS)

Kotamäki, M.; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

2016-01-01

Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

Progress in the integration of the ITER plant systems in auxiliary buildings

Energy Technology Data Exchange (ETDEWEB)

Kotamäki, M., E-mail: miikka.kotamaki@iter.org; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

2016-11-01

Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

Energy confinement scaling in tokamaks: some implications of recent experiments with ohmic and strong auxiliary heating

International Nuclear Information System (INIS)

Goldston, R.J.

1984-02-01

Recent results from confinement scaling experiments on tokamaks with ohmic and strong auxiliary heating are reviewed. An attempt is made to draw these results together into a low-density ohmic confinement scaling law, and a scaling law for confinement with auxiliary heating. The auxiliary heating confinement law may also serve to explain the saturation in tau/sub E/ vs anti n/sub e/ observed in some ohmic heating density scaling experiments

46 CFR 52.01-35 - Auxiliary, donkey, fired thermal fluid heater, and heating boilers.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Auxiliary, donkey, fired thermal fluid heater, and... (CONTINUED) MARINE ENGINEERING POWER BOILERS General Requirements § 52.01-35 Auxiliary, donkey, fired thermal... requirements for miscellaneous boiler types, such as donkey, fired thermal fluid heater, heating boiler, etc...

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

Intrinsically safe electrical installations, auxiliary circuits and electric communication equipment

Energy Technology Data Exchange (ETDEWEB)

Herms, C D

1981-11-19

Technical progress has not stopped short of electrical systems in mining, so that three new chapters are new included in the VDE regulations leaflet No. 0118 on 'Installation of electrical systems in underground coal mining'. The regulations on intrinsically safe electric systems, auxiliary circuits and communication systems are briefly described, and grounds for the regulations are presented. The regulations already take account of European regulations on intrinsic safety which will soon be published in a European Regulation on Mine Explosions. In the chapters on auxiliary circuits and communication systems, protection against direct contact, fires, and explosions is discussed as well as the further goal of reliable signal transmission.

Analysis and Measurement of NOx Emissions in Port Auxiliary Vessels

Directory of Open Access Journals (Sweden)

German de Melo Rodriguez

2013-09-01

Full Text Available This paper is made NOx pollution emitted by port auxiliary vessels, specifically by harbour tugs, due to its unique operating characteristics of operation, require a large propulsion power changes discontinuously, also possess some peculiar technical characteristics, large tonnage and high propulsive power, that differentiate them from other auxiliary vessels of the port. Taking into account all the above features, there are no studies of the NOx emission engines caused by different working regimes of power because engine manufacturers have not measured these emissions across the range of operating power, but usually we only report the pollution produced by its engines to a maximum continuous power.

29 CFR Appendix II to Part 1918 - Tables for Selected Miscellaneous Auxiliary Gear (Mandatory)

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Tables for Selected Miscellaneous Auxiliary Gear (Mandatory) II Appendix II to Part 1918 Labor Regulations Relating to Labor (Continued) OCCUPATIONAL SAFETY AND.... 1918, App. II Appendix II to Part 1918—Tables for Selected Miscellaneous Auxiliary Gear (Mandatory...

Performance of ALMR passive decay heat removal system

International Nuclear Information System (INIS)

Boardman, C.E.; Hunsbedt, A.

1991-01-01

The Advanced Liquid Metal Reactor (ALMR) concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the small (471 MWt) modular reactor to the environmental air by natural convection heat transfer. The system has no active components, requires no operator action to initiate, and is inherently reliable. The RVACS can perform its function under off-normal or degraded operating conditions without significant loss in performance. Several such events are described and the RVACS thermal performance for each is given and compared to the normal operation performance. The basic RVACS performance as well as the performance during several off-normal events have been updated to reflect design changes for recycled fuel with minor actinides for end of equilibrium cycle conditions. The performance results for several other off-normal events involving various degrees of RVACS air flow passage blockages are presented. The results demonstrated that the RVACS is unusually tolerant to a wide range of postulated faults. (author)

Adaptation of the MAST passive current simulation model for real-time plasma control

International Nuclear Information System (INIS)

McArdle, G.J.; Taylor, D.

2008-01-01

Successful equilibrium reconstruction on MAST depends on a reliable estimate of the passive current induced in the thick vacuum vessel (which also acts as the load assembly) and other toroidally continuous internal support structures. For the EFIT reconstruction code, a pre-processing program takes the measured plasma and PF coil current evolution and uses a sectional model of the passive structure to solve the ODEs for electromagnetic induction. The results are written to a file, which is treated by EFIT as a set of virtual measurements of the passive current in each section. However, when a real-time version of EFIT was recently installed in the MAST plasma control system, a similar function was required for real-time estimation of the instantaneous passive current. This required several adaptation steps for the induction model to reduce the computational overhead to the absolute minimum, whilst preserving accuracy of the result. These include: ·conversion of the ODE to use an auxiliary variable, avoiding the need to calculate the time derivative of current; ·minimise the order of the system via model reduction techniques with a state-space representation of the problem; ·transformation to eigenmode form, to diagonalise the main matrix for faster computation; ·discretisation of the ODE; ·hand-optimisation to use vector instruction extensions in the real-time processor; ·splitting the task into two parts: the time-critical feedback part, and the next cycle pre-calculation part. After these optimisations, the algorithm was successfully implemented at a cost of just 65 μs per 500 μs control cycle, with only 27 μs added to the control latency. The results show good agreement with the original off-line version. Some of these optimisations have also been used subsequently to improve the performance of the off-line version

Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

International Nuclear Information System (INIS)

Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

2003-01-01

The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

Large gap plasma display cell with auxiliary electrodes: macro-cell experiments and two-dimensional modelling

International Nuclear Information System (INIS)

Ouyang, J T; Callegari, Th; Caillier, B; Boeuf, J-P

2003-01-01

In this paper we use a two-dimensional fluid model and a 'macroscopic' PDP cell to investigate the possibility of using large gap configurations with auxiliary electrodes to improve the efficiency of PDP discharge cells. The large gap allows operation in a transient positive column regime where energy is more efficiently deposited into xenon excitation, while the auxiliary electrodes are used to keep reasonable values of the operating voltage. Two types of auxiliary electrode configurations (floating and powered) are considered. The discharge characteristics and the discharge efficiency in exciting xenon are studied with simulations and by measuring the intensity of infrared emission from xenon and visible emission from neon in a macroscopic PDP cell. The results show that an efficient positive column regime can be achieved at reasonably low operating voltages when the auxiliary electrode configuration is carefully designed

Site-specific analysis of hybrid geothermal/fossil power plants

Energy Technology Data Exchange (ETDEWEB)

1977-06-01

A preliminary economic analysis of a hybrid geothermal/coal power plant was completed for four geothermal resource areas: Roosevelt Hot Springs, Coso Hot Springs, East Mesa, and Long Valley. A hybrid plant would be economically viable at Roosevelt Hot Springs and somewhat less so at Coso Hot Springs. East Mesa and Long Valley show no economic promise. A well-designed hybrid plant could use geothermal energy for boiler feedwater heating, auxiliary power, auxiliary heating, and cooling water. Construction and operation of a hybrid plant at either Roosevelt Hot Springs or Coso Hot Springs is recommended. A modified version of the Lawrence Berkeley Livermore GEOTHM Program is the major analytical tool used in the analysis. The Intermountain Power Project is the reference all coal-fired plant.

Verb and auxiliary movement in agrammatic Broca's aphasia

NARCIS (Netherlands)

Bastiaanse, Y.R.M.; Thompson, C.K.

Verb production in agrammatic Broca's aphasia has repeatedly been shown to be impaired by a number of investigators. Not only is the number of verbs produced often significantly reduced, but verb inflections and auxiliaries are often omitted as well (e.g., Bastiaanse, Jonkers, & Moltmaker-Osinga,

Auxiliary Heat Exchanger Flow Distribution Test

International Nuclear Information System (INIS)

Kaufman, J.S.; Bressler, M.M.

1983-01-01

The Auxiliary Heat Exchanger Flow Distribution Test was the first part of a test program to develop a water-cooled (tube-side), compact heat exchanger for removing heat from the circulating gas in a high-temperature gas-cooled reactor (HTGR). Measurements of velocity and pressure were made with various shell side inlet and outlet configurations. A flow configuration was developed which provides acceptable velocity distribution throughout the heat exchanger without adding excessive pressure drop

Natural convection in an asymmetrically heated vertical channel with an adiabatic auxiliary plate

International Nuclear Information System (INIS)

Taieb, Soumaya; Hatem, Laatar Ali; Balti, Jalloul

2013-01-01

The effect of an auxiliary plate on natural convection in an asymmetrically heated channel is studied numerically in laminar regime. The computational procedure is made by solving the unsteady two dimensional Navier-Stokes and energy equations. This nonlinear system is integrated by a finite volume approach and then solved in time using the projection method, allowing the decoupling pressure from velocity. More than hundred simulations are performed to determine the best positions of the auxiliary plate that enhance the induced mass flow and the heat transfer rate for modified Rayleigh numbers ranging from Ra m = 10 2 to Ra m = 10 5 . Contour maps are plotted and then used to precise the enhancement rates of the mass flow and the heat transfer for any position of the auxiliary plate in the channel. The numerical results (velocity, pressure and temperature fields) provide detailed information about the evolution of the flow structure according to the geometry considered in this study. In addition, they permit to explain why the mass flow rate and Nusselt number are enhanced for certain positions of the auxiliary plate and are on the contrary deteriorated for others. (authors)

Total synthesis of cytochrome b562 by native chemical ligation using a removable auxiliary

Science.gov (United States)

Low, Donald W.; Hill, Michael G.; Carrasco, Michael R.; Kent, Stephen B. H.; Botti, Paolo

2001-01-01

We have completed the total chemical synthesis of cytochrome b562 and an axial ligand analogue, [SeMet7]cyt b562, by thioester-mediated chemical ligation of unprotected peptide segments. A novel auxiliary-mediated native chemical ligation that enables peptide ligation to be applied to protein sequences lacking cysteine was used. A cleavable thiol-containing auxiliary group, 1-phenyl-2-mercaptoethyl, was added to the α-amino group of one peptide segment to facilitate amide bond-forming ligation. The amine-linked 1-phenyl-2-mercaptoethyl auxiliary was stable to anhydrous hydrogen fluoride used to cleave and deprotect peptides after solid-phase peptide synthesis. Following native chemical ligation with a thioester-containing segment, the auxiliary group was cleanly removed from the newly formed amide bond by treatment with anhydrous hydrogen fluoride, yielding a full-length unmodified polypeptide product. The resulting polypeptide was reconstituted with heme and folded to form the functional protein molecule. Synthetic wild-type cyt b562 exhibited spectroscopic and electrochemical properties identical to the recombinant protein, whereas the engineered [SeMet7]cyt b562 analogue protein was spectroscopically and functionally distinct, with a reduction potential shifted by ≈45 mV. The use of the 1-phenyl-2-mercaptoethyl removable auxiliary reported here will greatly expand the applicability of total protein synthesis by native chemical ligation of unprotected peptide segments. PMID:11390992

Numerical analysis of magnetically suspended rotor in HTR-10 helium circulator being dropped into auxiliary bearings

International Nuclear Information System (INIS)

Zhao Jingxiong; Yang Guojun; Li Yue; Yu Suyuan

2012-01-01

Active magnetic bearings (AMB) have been selected to support the rotor of primary helium circulator in commercial 10 Mega-Walt High Temperature Gas-cooled Reactor (HTR-10). In an AMB system, the auxiliary bearings are necessary to protect the AMB components in case of losing power. This paper performs the impact simulation of Magnetically Suspended Rotor in HTR-10 Helium Circulator being dropped into the auxiliary bearings using the finite element program ABAQUS. The dynamic response and the strain field of auxiliary bearings are analyzed. The results achieved by the numerical analysis are in agreement with the experiment results. Therefore, the feasibility of the design of auxiliary bearing and the possibility of using the AMB system in the HTR are proved. (authors)

ASCERTAINMENT OF ELECTRIC-SUPPLY SCHEMES RELIABILITY FOR THE ATOMIC POWER PLANT AUXILIARIES

Directory of Open Access Journals (Sweden)

A. L. Starzhinskij

2015-01-01

Full Text Available The paper completes ascertainment of electrical-supply scheme reliability for the auxiliaries of a nuclear power plant. Thereat the author considers the system behavior during the block normal operation, carrying out current maintenance, and capital repairs in combination with initiating events. The initiating events for reactors include complete blackout, i.e. the loss of outside power supply (normal and reserve; emergency switching one of the working turbogenerators; momentary dumping the normal rating to the level of auxiliaries with seating the cutout valve of one turbo-generator. The combination of any initiating event with the repairing mode in case of one of the system elements failure should not lead to blackout occurrence of more than one system of the reliable power supply. This requirement rests content with the help of the reliable power supply system self-dependence (electrical and functional and the emergency power-supply operational autonomy (diesel generator and accumulator batteries.The reliability indicators of the power supply system for the nuclear power plant auxiliaries are the conditional probabilities of conjoined blackout of one, two, and three sections of the reliable power supply conditional upon an initiating event emerging and the blackout of one, two, and three reliable power-supply sections under the normal operational mode. Furthermore, they also are the blackout periodicity of one and conjointly two, three, and four sections of normal operation under the block normal operational mode. It is established that the blackout of one bus section of normal operation and one section of reliable power-supply system of the auxiliaries that does not lead to complete blackout of the plant auxiliaries may occur once in three years. The probability of simultaneous power failure of two or three normal-operation sections and of two reliable power-supply sections during the power plant service life is unlikely.

Component Data Base for Space Station Resistojet Auxiliary Propulsion

Science.gov (United States)

Bader, Clayton H.

1988-01-01

The resistojet was baselined for Space Station auxiliary propulsion because of its operational versatility, efficiency, and durability. This report was conceived as a guide to designers and planners of the Space Station auxiliary propulsion system. It is directed to the low thrust resistojet concept, though it should have application to other station concepts or systems such as the Environmental Control and Life Support System (ECLSS), Manufacturing and Technology Laboratory (MTL), and the Waste Fluid Management System (WFMS). The information will likely be quite useful in the same capacity for other non-Space Station systems including satellite, freeflyers, explorers, and maneuvering vehicles. The report is a catalog of the most useful information for the most significant feed system components and is organized for the greatest convenience of the user.

Integrated hydrogen control solutions for severe accidents using passive autocatalytic recombiners

International Nuclear Information System (INIS)

Bauer, M.; Tietsch, W.; Sabate Farnos, R.

2012-01-01

In a severe accident or a beyond-design-basis-accident, the reaction of water with zirconium alloy cladding, radiolysis of water, corium-concrete reactions and other corrosion phenomena generate hydrogen (H2). The detonation of this H2 in containment or in auxiliary buildings can result in damage to structures or loss of containment integrity. Identifying the generation and special distribution of hydrogen and controlling its concentration with Passive Autocatalytic Recombiners (PARs) solves this concern. Westinghouse's approach for hydrogen management starts by defining the quantities and transport/distribution of H 2 in-containment and out of containment with analysis tools such as MAAP, MELCOR, GASFLOW or FATE. Based on the results of these analyses, an optimized H2 Control Strategy is proposed in terms of number and location of PARs, and efficient integration with other H 2 management devices like e.g. existing igniters, H 2 monitors, etc.

Passive solar technology

Energy Technology Data Exchange (ETDEWEB)

Watson, D

1981-04-01

The present status of passive solar technology is summarized, including passive solar heating, cooling and daylighting. The key roles of the passive solar system designer and of innovation in the building industry are described. After definitions of passive design and a summary of passive design principles are given, performance and costs of passive solar technology are discussed. Passive energy design concepts or methods are then considered in the context of the overall process by which building decisions are made to achieve the integration of new techniques into conventional design. (LEW).

Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator; Modelado y simulacion del sistema de agua de alimentacion, controlador asociado e interfaz con el usuario para el simulador universitario de nucleoelectricas SUN-RAH

Energy Technology Data Exchange (ETDEWEB)

Sanchez B, A. [Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: alitet@eresmas.com

2003-07-01

The simulation process of the component systems of the feedwater of a nucleo electric plant is presented, using several models of reduced order that represent the diverse elements that compose the systems like: the heaters of feedwater, the condenser, the feedwater pump, etc. The integration of the same ones in one simulative structure, and the development of a platform that to give the appearance of to be executed in continuous time, it is the objective of the feedwater simulator, as well as of the SUN-RAH simulator, of which is part. The simulator uses models of reduced order that respond to the observed behavior of a nuclear plant of BWR type. Likewise, it is presented a model of a flow controller of feedwater that will be the one in charge of regulating the demand of the system according to the characteristics and criticize restrictions of safety and controllability, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. The integration of these models, the adaptation of the variables and parameters, are presented in a way that the integration with the other ones models of the remaining systems of the plant (reactor, steam lines, turbine, etc.), be direct and coherent with the principles of thermodynamic cycles relative to this type of generation plants. The design of those graphic interfaces and the environment where the simulator works its are part of those developments of this work. The reaches and objectives of the simulator complement the description of the simulator. (Author)

The CFS-PML for 2D Auxiliary Differential Equation FDTD Method Using Associated Hermite Orthogonal Functions

Directory of Open Access Journals (Sweden)

Feng Jiang

2017-01-01

Full Text Available The complex frequency shifted (CFS perfectly matched layer (PML is proposed for the two-dimensional auxiliary differential equation (ADE finite-difference time-domain (FDTD method combined with Associated Hermite (AH orthogonal functions. According to the property of constitutive parameters of CFS-PML (CPML absorbing boundary conditions (ABCs, the auxiliary differential variables are introduced. And one relationship between field components and auxiliary differential variables is derived. Substituting auxiliary differential variables into CPML ABCs, the other relationship between field components and auxiliary differential variables is derived. Then the matrix equations are obtained, which can be unified with Berenger’s PML (BPML and free space. The electric field expansion coefficients can thus be obtained, respectively. In order to validate the efficiency of the proposed method, one example of wave propagation in two-dimensional free space is calculated using BPML, UPML, and CPML. Moreover, the absorbing effectiveness of the BPML, UPML, and CPML is discussed in a two-dimensional (2D case, and the numerical simulations verify the accuracy and efficiency of the proposed method.

Generic description of facilities at the shaft head (auxiliary entrance installations) of deep geological repositories

International Nuclear Information System (INIS)

2016-10-01

In a deep geological repository, the access structures function as the link between the surface and the installations and structures at the disposal level. In the planned implementation scenarios, at least two access structures will be in operation up to the time of closure of the repository. The radioactive waste will be transported via the main access from the surface to the disposal level during emplacement operations. For the construction and operation of a deep geological repository, additional access structures are required. These auxiliary accesses and the associated surface infrastructure (e.g. shaft head installations) form the subject of this report. To provide as broad and comprehensive a description as possible, seven types of auxiliary access facilities are defined; these are characterised in line with the current status of planning and their functions and impacts are described. During construction, operation and dismantling of auxiliary access facilities, the usual conventional safety measures (inter alia) have to be observed (e.g. groundwater protection, fire prevention, facility security, accident prevention). Regarding the 'Ordinance on Protection against Major Accidents' no large quantities of hazardous materials, i.e. above the corresponding threshold quantities, are to be expected in the auxiliary access facilities. Proper handling and compliance with applicable regulations in all phases will ensure no hazard to humans and the environment. As no handling of radioactive materials is foreseen in the auxiliary access facilities, and because exhaust air and waste water from the controlled zones of a repository will, in principle, be removed via the main access and not the auxiliary accesses, a safety-relevant emission of radioactive substances and transport of contaminated material can be ruled out for the auxiliary access facilities during both normal operation and also in the case of an accident. Based on the information presented in

Study by the disco method of critical components of a P.W.R. normal feedwater system

International Nuclear Information System (INIS)

Duchemin, B.; Villeneuve, M.J. de; Vallette, F.; Bruna, J.G.

1983-03-01

The DISCO (Determination of Importance Sensitivity of COmponents) method objectif is to rank the components of a system in order to obtain the most important ones versus availability. This method uses the fault tree description of the system and the cut set technique. It ranks the components by ordering the importances attributed to each one. The DISCO method was applied to the study of the 900 MWe P.W.R. normal feedwater system with insufficient flow in steam generator. In order to take account of operating experience several data banks were used and the results compared. This study allowed to determine the most critical component (the turbo-pumps) and to propose and quantify modifications of the system in order to improve its availability

Vanishing auxiliary variables in PPS sampling - with applications in microscopy

DEFF Research Database (Denmark)

Andersen, Ina Trolle; Hahn, Ute; Jensen, Eva B. Vedel

Recently, non-uniform sampling has been suggested in microscopy to increase efficiency. More precisely, sampling proportional to size (PPS) has been introduced where the probability of sampling a unit in the population is proportional to the value of an auxiliary variable. Unfortunately, vanishing...... auxiliary variables are a common phenomenon in microscopy and, accordingly, part of the population is not accessible, using PPS sampling. We propose a modification of the design, for which an optimal solution can be found, using a model assisted approach. The optimal design has independent interest...... in sampling theory. We verify robustness of the new approach by numerical results, and we use real data to illustrate the applicability....

[The surgical nurse: his/her leadership of auxiliary nursing personnel].

Science.gov (United States)

Galvão, C M; Trevizan, M A; Sawada, N O; Mendes, I A

1997-01-01

This investigation as carried out in order to promote follow-up in the studies concerning nurse's leadership in the hospital context. Emphasys is given to the nurses that works in surgical ward unities. As a theoretical framework, authors utilized the model of leadership proposed by Hersey and Blanchard, named Situational Leadership. The objective was to analyze the correspondence of opinion between nurses and nursing auxiliary personnel about the leadership style of nurse should adopt in accordance with the maturity level of an element of the auxiliary personnel based on six categories of activities that were studied. Authors found out that nurses should adopt the styles of participant leadership, such as E3 (participating) and/or E4 (delegating).

A new auxiliary equation and exact travelling wave solutions of nonlinear equations

International Nuclear Information System (INIS)

Sirendaoreji

2006-01-01

A new auxiliary ordinary differential equation and its solutions are used for constructing exact travelling wave solutions of nonlinear partial differential equations in a unified way. The main idea of this method is to take full advantage of the auxiliary equation which has more new exact solutions. More new exact travelling wave solutions are obtained for the quadratic nonlinear Klein-Gordon equation, the combined KdV and mKdV equation, the sine-Gordon equation and the Whitham-Broer-Kaup equations

Performance assessment of auxiliary bearing in HTR-10 AMB helium circulator on the event of rotor drop

International Nuclear Information System (INIS)

Xiao Zhen; Yang Guojun; Li Yue; Shi Zhengang; Yu Suyuan

2014-01-01

In this paper, a model for analyzing internal contact stress arid external load of ball bearing from rotor displacement was developed based on the Hertz contact theory and applied to the analysis of the rotor drop test in HTR-10 helium circulator equipped with AMB (Active Magnetic Bearing) to gain a better understanding of auxiliary bearing performance at different stages after the rotor drop. It was shown that the auxiliary bearing can well resist axial impact produced by rotor drop, avoiding of internal severe plastic deformation and damage to the performance of the auxiliary bearing. Rotor's rotary motion and the heat accumulation of the inner ring resulted from the initial acute acceleration are the main contributor of radial load during the rotor idling and may cause the failure of auxiliary bearing. This paper analyzed the influence of this load and confirmed that the auxiliary bearing can still work in its loading limits. (authors)

Analysis for Passive Safety Injection of IPSS in Various LOCAs

International Nuclear Information System (INIS)

Kim, Sangho; Chang, Soonheung

2013-01-01

The Fukushima accident shows US the possibility of accidents that are beyond a designed imagination. Lots of lessons can be shortly summarized into three issues. First of all, the original cause was the occurrence of a Station Black-Out (SBO). Even if engineers considered the possibility of a loss of offsite power enough to be managed, the failure of EDGs seemed to be unnoticed. The second is poor operation and accident management. They could not understand the overall system and did not check the availability of alternating systems. The third is the large release of radioactive materials outside the containment. Even if SBO occurred and the accident was not managed well, all the means must have prevented the large release out of containment. After that, lots of problems were pointed and numerous actions were carried out in each country. The representative proposals are AAC, additional physical barrier, bunker concept and large big tank. Integrated passive safety system (IPSS) was proposed as one of the solutions for enhancing the safety. IPSS can cope with a SBO and accidents with a SBO. IPSS has five functions which are passive decay heat removal, passive safety injection, passive containment cooling, passive in-vessel retention and filtered venting system. The results showed a high performance of removing decay heat through steam generator cooling by forming natural circulation in the primary circuit. The design concept of passive safety injection system (PSIS) consists of the injection line from integrated passive safety tank (IPST) to reactor vessel. The previous works were only focused on a double ended guillotine break LOCA in SBO. The purpose of this paper is to analyze the performance of PSIS in IPSS for various LOCAs by using MARS (Multi-dimensional Analysis of Reactor Safety) code. The simulated accidents were LOCAs which were accompanied with a SBO. The conditions of the LOCAs were varied only for the size of break. It shall show the capability of PSIS

Auxiliary fields for super Yang-Mills from division algebras

CERN Document Server

Evans, Jonathan M.

1994-01-01

Division algebras are used to explain the existence and symmetries of various sets of auxiliary fields for super Yang-Mills in dimensions d=3,4,6,10. (Contribution to G\\"ursey Memorial Conference I: Strings and Symmetries)

Use of the GO methodology to assess plant operating data

International Nuclear Information System (INIS)

Wenger, L.O.; Danner, M.A.

1987-01-01

In September 1985, the Institute of Nuclear Power Operations (INPO) published The Operational Performance of Auxiliary Feedwater Systems in US PWRs, 1980-1984, in which many of the problems experienced by the nuclear industry with auxiliary feedwater systems (AFWSs) were summarized. The principal conclusions of the INPO report were that pressurized water reactor AFWS equipment failure rates had nearly doubled industry-wide since 1980. Calvert Cliffs Units 1 and 2 were highlighted by INPO as being among some of the poorer performers identified in the report. The study objective involved estimating the historical AFWS unavailability so that it could be compared with a predetermined acceptable availability goal. From this comparison, the relative importance of individual contributions to overall system unavailability resulting from various equipment failure modes could be assessed. The GO methodology, a computerized reliability analyses technique, was selected as the means for performing additional analyses. A GO system model was developed to represent the current Calvert Cliffs AFWS configuration. In summary, it was shown that integration of plant data with a comprehensive availability/reliability model can provide both the necessary insights to focus attention on the most important contributions to system unavailability and a pragmatic way for developing a cost effective implementation strategy. The GO methodology was useful and provided the requisite results with a minimal outlay of valuable resources

Analysis of loss of offsite power transient using RELAP5/MOD1/NSC

International Nuclear Information System (INIS)

Kim, Hho Jung; Chung, Bub Dong; Lee, Young Jin; Kim, Jin Soo

1986-01-01

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAP5/MOD1/NSC), based upon the sequence of events for the KNU1( Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximate the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV setting etc. are recognized to be important in the transient analyses on a bestestimate basis. (Author)

ADAPTIVE SELECTION OF AUXILIARY OBJECTIVES IN MULTIOBJECTIVE EVOLUTIONARY ALGORITHMS

Directory of Open Access Journals (Sweden)

I. A. Petrova

2016-05-01

Full Text Available Subject of Research.We propose to modify the EA+RL method, which increases efficiency of evolutionary algorithms by means of auxiliary objectives. The proposed modification is compared to the existing objective selection methods on the example of travelling salesman problem. Method. In the EA+RL method a reinforcement learning algorithm is used to select an objective – the target objective or one of the auxiliary objectives – at each iteration of the single-objective evolutionary algorithm.The proposed modification of the EA+RL method adopts this approach for the usage with a multiobjective evolutionary algorithm. As opposed to theEA+RL method, in this modification one of the auxiliary objectives is selected by reinforcement learning and optimized together with the target objective at each step of the multiobjective evolutionary algorithm. Main Results.The proposed modification of the EA+RL method was compared to the existing objective selection methods on the example of travelling salesman problem. In the EA+RL method and its proposed modification reinforcement learning algorithms for stationary and non-stationary environment were used. The proposed modification of the EA+RL method applied with reinforcement learning for non-stationary environment outperformed the considered objective selection algorithms on the most problem instances. Practical Significance. The proposed approach increases efficiency of evolutionary algorithms, which may be used for solving discrete NP-hard optimization problems. They are, in particular, combinatorial path search problems and scheduling problems.

Multireference Density Functional Theory with Generalized Auxiliary Systems for Ground and Excited States.

Science.gov (United States)

Chen, Zehua; Zhang, Du; Jin, Ye; Yang, Yang; Su, Neil Qiang; Yang, Weitao

2017-09-21

To describe static correlation, we develop a new approach to density functional theory (DFT), which uses a generalized auxiliary system that is of a different symmetry, such as particle number or spin, from that of the physical system. The total energy of the physical system consists of two parts: the energy of the auxiliary system, which is determined with a chosen density functional approximation (DFA), and the excitation energy from an approximate linear response theory that restores the symmetry to that of the physical system, thus rigorously leading to a multideterminant description of the physical system. The electron density of the physical system is different from that of the auxiliary system and is uniquely determined from the functional derivative of the total energy with respect to the external potential. Our energy functional is thus an implicit functional of the physical system density, but an explicit functional of the auxiliary system density. We show that the total energy minimum and stationary states, describing the ground and excited states of the physical system, can be obtained by a self-consistent optimization with respect to the explicit variable, the generalized Kohn-Sham noninteracting density matrix. We have developed the generalized optimized effective potential method for the self-consistent optimization. Among options of the auxiliary system and the associated linear response theory, reformulated versions of the particle-particle random phase approximation (pp-RPA) and the spin-flip time-dependent density functional theory (SF-TDDFT) are selected for illustration of principle. Numerical results show that our multireference DFT successfully describes static correlation in bond dissociation and double bond rotation.

Controlling indoor climate. Passive cooling of residential buildings in hot-humid climates in China

Energy Technology Data Exchange (ETDEWEB)

Zhiwu, Wang

1996-10-01

Overheating is a paramount problem in residential buildings in hot and humid climates in China during summer. This study aims to deal with the overheating problem and the problem of poor air quality in dwellings. The main objective is to improve indoor thermal conditions by passive cooling approaches, climatisation techniques in buildings without auxiliary cooling from air conditioning equipment. This thesis focuses on the study of cross-ventilation in apartments, which is one of the most effective ways of natural cooling in a hot humid climate, but is also one of the least understood parts in controlling indoor climate. The Computational Fluid Dynamics (CFD) technique is used, which is a new approach, since cross-ventilation studies have been conventionally made by wind tunnel tests. The validations of the CFD technique are examined by a comparison between wind tunnel tests and computer simulations. The factors influencing indoor air movement are investigated for a single room. Cross-ventilation in two apartments is studied, and the air change efficiency in a Chinese kitchen is calculated with CFD techniques. The thermal performance of ventilated roofs, a simple and widely used type of roof in the region, is specially addressed by means of a full-scale measurement, wind tunnel tests and computer simulations. An integrated study of passive cooling approaches and factors affecting indoor thermal comfort is carried out through a case study in a southern Chinese city, Guangzhou. This thesis demonstrates that passive cooling measure have a high potential in significantly improving indoor thermal conditions during summer. This study also gives discussions and conclusions on the evaluation of indoor thermal environment; effects influencing cross-ventilation in apartments; design guidelines for ventilated roofs and an integrated study of passive cooling. 111 refs, 83 figs, 65 tabs

Model predictions for auxiliary heating in spheromaks

International Nuclear Information System (INIS)

Fauler, T.K.; Khua, D.D.

1997-01-01

Calculations are presented of the plasma temperature waited for under auxiliary heating in spheromaks. A model, ensuring good agreement of earlier experiments with joule heating results, is used. The model includes heat losses due to magnetic fluctuations and shows that the plasma temperatures of the kilo-electron-volt order may be achieved in a small device with the radius of 0.3 m only

A Corpus-Based Study on the Use of Past Tense Auxiliary "Be" in Argumentative Essays of Malaysian ESL Learners

Science.gov (United States)

Manokaran, Janaki; Ramalingam, Chithra; Adriana, Karen

2013-01-01

This research is a corpus-based study of secondary and college ESL Malaysian learner's written work by identifying and classifying the types of errors in the Past Tense Auxiliary "Be". This research studied the past tense auxiliary "be", types of past tense auxiliary "be" errors and frequency of past tense auxiliary…

Versatile Auxiliary Orthodontic Spring for Orthodontic Correction of Impacted Teeth

Directory of Open Access Journals (Sweden)

Pavankumar Janardan Vibhute

2011-01-01

Full Text Available Malocclusion such as impacted tooth is not uncommon. Many approaches with various auxiliary springs have been reported in literature till date for correction of such malocclusions. They had biomechanical, retentive and stability drawbacks inherent in their designs. This article presents the innovative approach for orthodontic correction of impacted tooth, especially with light force appliance, i.e. Begg′s appliance, where round wires in round molar tubes are used throughout treatment. A versatile auxiliary orthodontic spring (VAOS is fabricated in the 0.018 inch Australian stainless steel round wire, which may be anchored on round molar buccal tube, and desirable force vector may be applied in any of the three dimensions. Fabrication and its clinical application are discussed.

Data report for ROSA-IV LSTF 10% hot leg break experiment Run SB-HL-02

International Nuclear Information System (INIS)

Kukita, Yutaka; Hirata, Kazuo; Gotou, Hiroki

1990-03-01

Experimental data for the 10% hot leg break test, Run SB-HL-02, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on June 30, 1987, are presented. This test assumed total failure of both high pressure injection (HPI) and auxiliary feedwater (AFW) systems. The test results were characterized by asymmetric loop responses, flashing in the cold legs and upper downcomer, as well as condensation depressurization in the cold legs following injection of emergency core coolant (ECC) from accumulators. (author)

A new Zero-Voltage-Transition PWM switching cell

Energy Technology Data Exchange (ETDEWEB)

Grigore, V. [Electronics and Telecommunications Faculty `Politebuica` University Bucharest (Romania); Kyyrae, J. [Helsinki University of Technology, Otaniemi (Finland): Institute of Intelligent Power Electronics

1997-12-31

In this paper a new Zero-Voltage-Transition (ZVT) PWM switching cell is presented. The proposed switching cell is composed of the normal hard-switched PWM cell (consisting of one active switch and one passive switch), plus an auxiliary circuit (consisting of one active switch and some reactive components). The auxiliary circuit is inactive during the ON and OFF intervals of the switches in the normal PWM switch. However, the transitions between the two states are controlled by the auxiliary circuit. Prior to turn-on, the voltage across the active switch in the PWM cell is forced to zero, thus the turn-on losses of the active switch are practically eliminated. At turn-off the auxiliary circuit behaves like a non-dissipative passive snubber reducing the turn-off losses to a great extent. Zero-Voltage-Transition switching technique almost eliminates switching losses. The active switch operates under ZVT conditions, the passive switch (diode) has a controlled reverse recovery, and the switch in the auxiliary circuit operates under Zero-Current-Switching (ZCS) conditions. (orig.) 6 refs.

Reduction of residual gas in a sputtering system by auxiliary sputter of rare-earth metal

International Nuclear Information System (INIS)

Li Dejie

2002-01-01

In film deposition by sputtering, the oxidation and nitrification of the sputtered material lead to degradation of film quality, particularly with respect to metal sulfide films. We propose to use auxiliary sputtering as a method to produce a fresh film of rare-earth metal, usually dysprosium (Dy), that absorbs the active gases in a sputtering system, greatly reducing the background pressure and protecting the film from oxidation and nitrification effectively. The influence of the auxiliary sputtering power consumption, sputtering time, and medium gas pressure on the background pressure in the vacuum chamber is investigated in detail. If the auxiliary sputtering power exceeds 120 W and the sputtering time is more than 4 min, the background pressure is only one fourth of the ultimate pressure pumped by an oil diffusion pump. The absorption activity of the sputtered Dy film continues at least an hour after completion of the auxiliary sputter. Applied to film deposition of Ti and ZnS, this technique has been proven to be effective. For the Ti film, the total content of N and O is reduced from 45% to 20% when the auxiliary sputtering power of Dy is 120 W, and the sputtering time is 20 min. In the case of ZnS, the content of O is reduced from 8% to 2%

Meaningful questions: The acquisition of auxiliary inversion in a connectionist model of sentence production.

Science.gov (United States)

Fitz, Hartmut; Chang, Franklin

2017-09-01

Nativist theories have argued that language involves syntactic principles which are unlearnable from the input children receive. A paradigm case of these innate principles is the structure dependence of auxiliary inversion in complex polar questions (Chomsky, 1968, 1975, 1980). Computational approaches have focused on the properties of the input in explaining how children acquire these questions. In contrast, we argue that messages are structured in a way that supports structure dependence in syntax. We demonstrate this approach within a connectionist model of sentence production (Chang, 2009) which learned to generate a range of complex polar questions from a structured message without positive exemplars in the input. The model also generated different types of error in development that were similar in magnitude to those in children (e.g., auxiliary doubling, Ambridge, Rowland, & Pine, 2008; Crain & Nakayama, 1987). Through model comparisons we trace how meaning constraints and linguistic experience interact during the acquisition of auxiliary inversion. Our results suggest that auxiliary inversion rules in English can be acquired without innate syntactic principles, as long as it is assumed that speakers who ask complex questions express messages that are structured into multiple propositions. Copyright © 2017 Elsevier B.V. All rights reserved.

Reliability of supply of switchgear for auxiliary low voltage in substations extra high voltage to high voltage

Directory of Open Access Journals (Sweden)

Perić Dragoslav M.

2015-01-01

Full Text Available Switchgear for auxiliary low voltage in substations (SS of extra high voltages (EHV to high voltage (HV - SS EHV/HV kV/kV is of special interest for the functioning of these important SS, as it provides a supply for system of protection and other vital functions of SS. The article addresses several characteristic examples involving MV lines with varying degrees of independence of their supply, and the possible application of direct transformation EHV/LV through special voltage transformers. Auxiliary sources such as inverters and diesel generators, which have limited power and expensive energy, are also used for the supply of switchgear for auxiliary low voltage. Corresponding reliability indices are calculated for all examples including mean expected annual engagement of diesel generators. The applicability of certain solutions of switchgear for auxiliary low voltage SS EHV/HV, taking into account their reliability, feasibility and cost-effectiveness is analyzed too. In particular, the analysis of applications of direct transformation EHV/LV for supply of switchgear for auxiliary low voltage, for both new and existing SS EHV/HV.

Safety analysis of high temperature reactor cooled and moderated by supercritical light water

International Nuclear Information System (INIS)

Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

2003-01-01

This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

Generating Selected Color using RGB, Auxiliary Lights, and Simplex Search

Directory of Open Access Journals (Sweden)

Kim HyungTae

2015-01-01

Full Text Available A mixed light source generates various colors, with the potential to adjust intensities of multiple LEDs, which makes it possible to generate arbitrary colors. Currently, PCs and OSs provide color selection windows that can obtain the RGB or HSL color coordinates of a user’s selection. Mixed light sources are usually composed of LEDs in the primary colors, with LEDs in auxiliary colors such as white and yellow used in a few cases. When using auxiliary color LEDs, the number of LED inputs, the dimming levels, is larger than the number of elements in the color coordinate, which causes an under-determined problem. This study proposed how to determine the dimming levels of LEDs based on the selected color. Commercial LEDs have di_erent optical power values and impure color coordinates, even if they are RGB. Hence, the characteristics of the LEDs were described using a linear model derived from the tri-stimulus values (an XYZ color coordinate model and dimming levels. Color mixing models were derived for the arbitrary number of auxiliary color LEDs. The under-determined problem was solved using a simplex search method without an inverse matrix operation. The proposed method can be applied to a machine vision system and an RGBW light mixer for semiconductor inspection. The dimming levels, obtained using the proposed method were better than derived using other methods.

Study on primary coolant system depressurization effect factor in pressurized water reactor

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

Review of passive-blind detection in digital video forgery based on sensing and imaging techniques

Science.gov (United States)

Tao, Junjie; Jia, Lili; You, Ying

2016-01-01

Advances in digital video compression and IP communication technologies raised new issues and challenges concerning the integrity and authenticity of surveillance videos. It is so important that the system should ensure that once recorded, the video cannot be altered; ensuring the audit trail is intact for evidential purposes. This paper gives an overview of passive techniques of Digital Video Forensics which are based on intrinsic fingerprints inherent in digital surveillance videos. In this paper, we performed a thorough research of literatures relevant to video manipulation detection methods which accomplish blind authentications without referring to any auxiliary information. We presents review of various existing methods in literature, and much more work is needed to be done in this field of video forensics based on video data analysis and observation of the surveillance systems.

Pressurizer /Auxiliary Spray Piping Stress Analysis For Determination Of Lead Shielding Maximum Allow Able Load

International Nuclear Information System (INIS)

Setjo, Renaningsih

2000-01-01

Piping stress analysis for PZR/Auxiliary Spray Lines Nuclear Power Plant AV Unit I(PWR Type) has been carried out. The purpose of this analysis is to establish a maximum allowable load that is permitted at the time of need by placing lead shielding on the piping system on class 1 pipe, Pressurizer/Auxiliary Spray Lines (PZR/Aux.) Reactor Coolant Loop 1 and 4 for NPP AV Unit one in the mode 5 and 6 during outage. This analysis is intended to reduce the maximum amount of radiation dose for the operator during ISI ( In service Inspection) period.The result shown that the maximum allowable loads for 4 inches lines for PZR/Auxiliary Spray Lines is 123 lbs/feet

Cooling system for auxiliary systems of a nuclear power plant

International Nuclear Information System (INIS)

Maerker, W.; Mueller, K.; Roller, W.

1981-01-01

From the reactor auxiliary and ancillary systems of a nuclear facility heat has to be removed without the hazard arising that radioactive liquids or gases may escape from the safe area of the nuclear facility. A cooling system is described allowing at every moment to make available cooling fluid at a temperature sufficiently low for heat exchangers to be able to remove the heat from such auxiliary systems without needing fresh water supply or water reservoirs. For this purpose a dry cooling tower is connected in series with a heat exchanger that is cooled on the secondary side by means of a refrigerating machine. The cooling pipes are filled with a nonfreezable fluid. By means of a bypass a minimum temperature is guaranteed at cold weather. (orig.) [de

[Evaluating work intensity in major and auxiliary occupations of by-product coke industry].

Science.gov (United States)

Smagulov, N K; Alpysbayeva, Zh T

2015-01-01

The article covers evaluation of work strain in major and auxiliary occupations of by-product coke industry. The study results conclude that occupational activity of by-product coke industry workers, under exposure to occupational hazards, affects the workers' performance. Major occupations workers demonstrate higher level of functional strain of CNS, poor concentration of attention and lower ability to switch over, decreased general performance, vs. the auxiliary occupations workers who demonstrated increased cardiovascular and neuro-muscular strain due to occupational activity.

Pure spinors as auxiliary fields in the ten-dimensional supersymmetric Yang-Mills theory

International Nuclear Information System (INIS)

Nilsson, B.E.W.

1986-01-01

A new way of introducing auxiliary fields into the ten-dimensional supersymmetric Yang-Mills theory is proposed. The auxiliary fields are commuting 'pure spinors' and constitute a non-linear realisation of the Lorentz group. This invalidates previous no-go theorems concerning the possibility of going off-shell in this theory. There seems to be a close relation between pure spinors and the concepts usually used in twistor theory. The non-Abelian theory can be constructed for all groups having pseudo-real representations. (author)