Travis, R.; Higgins, J.; Gunther, W.; Shier, W.
The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A system Risk-based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Susquehanna Steam Electric Station (SSES) which is operated by Pennsylvania Power ampersand Light (PP ampersand L). Included in this S-RIG is a discussion of the role of HPCI in mitigating accidents and a presentation of PRA-based failure modes which could prevent proper operation of the system. The S-RIG uses industry operating experience, including plant-specific illustrative examples, to augment the basic PRA failure modes. It is designed to be used as a reference for both routine inspections and the evaluation of the significance of component failures
Shier, W.; Gunther, W.
A review of the operating experience for the High Pressure Coolant Injection (HPCI) system at the Pilgrim Nuclear Power Station is described in this report. The information for this review was obtained from Pilgrim Licensee Event Reports (LERs) that were generated between 1980 and 1989. These LERs have been categorized into 23 failure modes that have been prioritized based on probabilistic risk assessment considerations. In addition, the results of the Pilgrim operating experience review have been compared with the results of of a similar, industry wide operating experience review. this comparison provides an indication of areas in the Pilgrim HPCI system that should be given increased attention in the prioritization of inspection resources
Wong, S.; DiBiasio, A.; Gunther, W.
The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant
Wong, S.; DiBiasio, A.; Gunther, W. [Brookhaven National Lab., Upton, NY (United States)
The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.
Christie, R.F.; Stetkar, J.W.
The change in availability of the high-pressure coolant injection system (HPCIS) due to a change in pump and valve test interval from monthly to quarterly was analyzed. This analysis started by using the HPCIS base line evaluation produced as part of the Browns Ferry Nuclear Plant (BFN) Probabilistic Risk Assessment (PRA). The base line evaluation showed that the dominant contributors to the unavailability of the HPCI system are hardware failures and the resultant downtime for unscheduled maintenance. The effect of changing the pump and valve test interval from monthly to quarterly was analyzed by considering the system unavailability due to hardware failures, the unavailability due to testing, and the unavailability due to human errors that potentially could occur during testing. The magnitude of the changes in unavailability affected by the change in test interval are discussed. The analysis showed a small increase in the availability of the HPCIS to respond to loss of coolant accidents (LOCAs) and a small decrease in the availability of the HPCIS to respond to transients which require HPCIS actuation. In summary, the increase in test interval from monthly to quarterly does not significantly impact the overall HPCIS availability
Bratfisch, Christoph; Koch, Marco K. [Ruhr-Univ. Bochum (Germany). Reactor Simulation and Safety Group
For extented application and analyses of the severe accident code ATHLET-CD, the course of the invessel accident in Unit 3 of Fukushima-Daiichi is simulated in the frame of the research project SUBA as a part of the BMBF sponsored collaborative project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Stoerfallmassnahmen). Investigations, carried out by TEPCO, had shown that the High-Pressure Coolant Injection system (HPCI) might have stopped earlier than expected. A parameter variation was performed to analyze the impact of the tripped HPCI injection regarding the thermohydraulic behaviour as well as the core degradation phenomena.
In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. It is estimated that the water injection by HPIC was stopped since around at 13/3 19:00 and the water level in RPV decreased since then. It is estimated that the RPV broke at 14/3 8:55 and water could injected from fire engines due to the depression due to the rupture of RPV. There was little water left in RPV at the time of the rupture. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. If HPCI was working properly until the operators stopped it, the plant parameters obtained in the accident cannot be explained. (author)
Matteoli, Sara; Wilhjelm, Jens E.; Torp-Pedersen, Soren T.
The human heel pad is a complex structure that features non-linear visco-elastic characteristics as the majority of the human soft tissues. The biomechanical aspects of the heel pad are still under investigation and the influence of subject factors such as age, weight, gender, height, race......, and body activity have been reported. The aim of this paper is to study the literature in order to identify the influence of subject factors and diseases on the heel pad compressibility index....
Kaulitz, D. E.
Most early vintage Boiling Water Reactors have a high head and high capacity High Pressure Coolant Injection (HPCI) pump to keep the core covered following a loss of coolant accident (LOCA). However, the protection afforded by the HPCI pump for mitigating a LOCA introduces the potential that a spurious start of the HPCI pump could oversupply the reactor vessel and lead to an automatic trip of the main turbine due to high water level. A turbine trip and associated increase in moderator density could challenge the bases of fuel integrity operating limits. To prevent turbine trip during spurious operation of the HPCI pump, the reactor protection system includes instrumentation and logic to sense high water level and automatically trip the HPCI pump prior to reaching the turbine trip setpoint. This paper describes an analysis that was performed to determine if existing reactor vessel water level trip instrumentation, logic and setpoints result in a high probability that the HPCI pump will trip prior to actuation of the turbine trip. Using nominal values for the initial water level and for the HPCI pump and turbine trip setpoints, and using the probability distribution functions for measurement uncertainty in these setpoints, a Monte Carlo simulation was employed to determine probabilities of successfully tripping the HPCI pump prior to tripping of the turbine. The results of the analysis established that the existing setpoints, instrumentation and logic would be expected to reliably prevent a trip of the main turbine. (authors)
through the monochromator at zero order) was incident on the sample and the corresponding current in the circuit noted using an electrometer (Keithley 614). The photocurrent at similar conditions of incident intensity is greater in HPCI than in H, PC ...
Rajan, Gunesh; Tavora-Vieira, Dayse; Baumgartner, Wolf-Dieter; Godey, Benoit; Müller, Joachim; O'Driscoll, Martin; Skarzynski, Henryk; Skarzynski, Piotr; Usami, Shin-Ichi; Adunka, Oliver; Agrawal, Sumit; Bruce, Iain; De Bodt, Marc; Caversaccio, Marco; Pilsbury, Harold; Gavilán, Javier; Hagen, Rudolf; Hagr, Abdulrahman; Kameswaran, Mohan; Karltorp, Eva; Kompis, Martin; Kuzovkov, Vlad; Lassaletta, Luis; Yongxin, Li; Lorens, Artur; Manoj, Manikoth; Martin, Jane; Mertens, Griet; Mlynski, Robert; Parnes, Lorne; Pulibalathingal, Sasidharan; Radeloff, Andreas; Raine, Christopher H; Rajeswaran, Ranjith; Schmutzhard, Joachim; Sprinzl, Georg; Staecker, Hinrich; Stephan, Kurt; Sugarova, Serafima; Zernotti, Mario; Zorowka, Patrick; Van de Heyning, Paul
To provide multidisciplinary cochlear implant teams with a current consensus statement to support hearing preservation cochlear implantation (HPCI) in children, including those children with symptomatic partial deafness (PD) where the intention is to use electric-acoustic stimulation (EAS). The main objectives are to provide guidelines on who is a candidate, how to assess these children and when to implant if Med-El Flex electrode arrays are chosen for implantation. The HEARRING group reviewed the current evidence and practice regarding the management of children to be considered for HPCI surgery emphasizing the assessment needed prior to implantation in order to demonstrate the benefits in these children over time. The consensus statement addresses following three key questions: (1) Should these children be treated? (2) How to identify these children? (3) How to manage these children? The HEARRING group concludes that irrespective of the degree of residual hearing present, the concepts of hearing and structure preservation should be applied in every child undergoing cochlear implantation and that HPCI is a safe and reliable treatment option. Early detection and multidisciplinary assessment are key to the identification of children with symptomatic PD, these children should undergo HPCI as early as possible.
A review of the operating experience for the High Pressure Coolant Injection (HPCI) system at the Hatch Nuclear Power Station, Units 1 and 2, is described in this report. The information for this review was obtained from Hatch Licensee Event Reports (LERs) that were generated between 1980 and 1992. These LERs have been categorized into 23 failure modes that have been prioritized based on probabilistic risk assessment considerations. In addition, the results of the Hatch operating experience review have been compared with the results of a similar, industry wide operating, experience review. This comparison provides an indication of areas in the Hatch HPCI system that should be given increased attention in the prioritization of inspection resources
On July 1, 1988, a high pressure coolant injection (HPCI) steam admission valve failed to open during a post-maintenance test at the Brunswick nuclear power plant, Unit 1. The same valve had failed in December 1987 and on May 28, 1988. The licensee, Carolina Power and Light Company, established a team to investigate the cause of failure, and the team identified the most probable cause as a dc motor failure due to a shunt-winding to series-winding short circuit. The team believed that this condition was precipitated by thermal binding of the valve internals. The previous failure in May was also diagnosed as having been caused by thermal binding. As a result of these failures, the licensee reviewed the design of the dc motor-operated valves for both the HPCI and the reactor core isolation cooling (RCIC) systems. This review identified a number of significant design deficiencies going well beyond the problems with thermal binding. The deficiencies constitute a potential common cause failure mechanism for safety system valves. Unit 1 was shut down on July 14, 1988 to replace the failed HPCI valve motor and to implement design modifications to other motor-operated valves
Lee, J.H.; Oka, Y.; Koshizuka, S.
The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics
The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case
The equipment qualification program described in this plan is intended to provide the technical basis for resolving uncertainties in existing equipment qualification standards. In addition, research results are contributing to the resolution of safety issues GI-23, GI-87, USI-A44, titled, ''Reactor Coolant Pump Seal Failure,'' ''Failure of HPCI Steam Line Without Isolation,'' and ''Station Blackout,'' respectively. Also, research effort is being directed at providing information on the behavior of containment isolation valves under severe accident environments. Although the results of the latter research will not contribute to resolving uncertainties in qualification standards, it has proven cost effective to obtain this information under this program
Government's Investigation Committee on the Accident at Fukushima Nuclear Power Stations of Tokyo Electric Power Company published its final report on July 23, 2012. Results of investigation combined final report and interim report published on December 26, 2011. The author was head of accident accuse investigation team mostly in charge of site response, prior measure and plant behavior. This article reported author related technical investigation results focusing on site response and prior measures against tsunamis of units 1-3 of Fukushima Nuclear Power Stations. Misunderstanding of working state of isolation condenser of unit 1, unsuitability of alternative water injection at manual stop of high-pressure coolant injection (HPCI) system of unit 3 and improper prior measure against tsunami and severe accident were pointed out in interim report. Improper monitoring of suppression chamber of unit 2 and again unsuitable work for HPCI system of unit 3 were reported in final report. Thorough technical investigation was more encouraged to update safety measures of nuclear power stations. (T. Tanaka)
Heising, C.D.; Dinsmore, S.C.
This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to construct and quantitatively evaluate an alarm analysis system for the engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide system failure mode information which can be used to construct alarm trees. These alarm trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel. In the alarm tree, possible states of each instrumented alarem are identified as true or false. In addition, a warning status is defined and integrated into an alarm analysis routine. The impact of this additional status conditioned on the Boolean laws used to evaluate the alarm trees is examined. An application is described for BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents
Results of the ROSA-II test simulating a loss-of coolant accident (LOCA) in a light water reactor (LWR) are presented, including the test conditions and interpretation of the phenomena for test runs 415, 417, 421 and 422. Even in small break at the cold leg, the core is exposed to void and the temperature rises. In small break of the hot leg, however, core cooling keeps without temperature rise, because there still remains much residual water and upward core flow exists. Direct effect of the HPCI on the depressurization rate is small, but it increases the accumulator injection rate, leading to early core reflooding and early core cooling from upward. Effects of the secondary system depressurization are increase of depressurization and discharge rates of the primary loop, which results in early initiation of the accumulator injection and core reflooding. (auth.)
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 stem turbine drive 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 Cooking (RCIC) and High Pressure Coolant Injection (HPCI) systems for Boiling Water Reactor (BWR) facilities. This research has been conducted by examining current information in the Nuclear Plant Reliability Data System (NPRDS), reviewing Licensee Event Reports, thoroughly investigating contacts with operating plant personnel, and by personal observation. This information was reviewed to determine the cause of each reported 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
Human Periapical Cysts-Mesenchymal Stem Cells Cultured with Allogenic Human Serum are a “clinical-grade” construct alternative to bovine fetal serum and indicated in the regeneration of endo-periodontal tissues
Full Text Available Aim: Our research investigated the use of human serum (HS as a safe and clinical-grade culture medium, using a new cell-model: hPCy-MSCs. This article is aimed to concretely applicate the concept of “waste-based regenerative dentistry” to translate it in future endo-periodontal applications. Methodology: HPCy-MSCs were cultured in 2 different mediums, both containing α-MEM: the 1st with 10% FBS (Control group, and the 2nd with 10% human serum (Test group.Cell proliferation and stemness assays, gene expression, immunophenotypic analysis and osteogenic differentiation were performed to verify our hypothesis. cDNA samples were amplified with qPCR.Experiments were performed in triplicate and analysed with statistical software. Results: The hPCy-MSCs cultivated in a medium with HS were morphologically similar to those cultivated with FBS, and showed a significantly higher proliferation rate. Von Kossa's staining revealed that osteoblasts from hPCy-MSCs in HS implemented with osteogenic induction factors, showed a better osteogenic activity, also confirmed by a significant upregulation of osteopotin (OPN and matrix extracellular phosphoglycoprotein (MEPE. Conclusions: HPCy-MSCs cultivated in HS showed phenotypic stability and a clear regenerative binding, thus, suggesting these two components as a clinically-grade construct for future endo-periodontal therapies. Riassunto: Obiettivi: La nostra ricerca ha analizzato l’utilizzo del siero umano (HS come mezzo di coltura sicuro e “clinical-grade”, per uso clinico, utilizzando un nuovo modello cellulare: le hPC-MSCs. Questo articolo ha lo scopo di applicare concretamente il concetto di “odontoiatria rigenerativa basata sui rifiuti biologici”, al fine di tradurlo in future applicazioni endo-periodontali. Materiali e metodi: Le HPCy-MSCs sono state coltivate in 2 mezzi di coltura diversi, entrambi contenenti α-MEM: il primo con 10% di FBS (gruppo di controllo e il secondo con il 10% di siero
Steele, R. Jr.; DeWall, K.G.
This paper presents preliminary observations from the US Nuclear Regulatory Commission/Idaho National Engineering Laboratory Flexible Wedge Gate Valve Qualification and Flow Interruption Test Program, Phase 2. The program investigated the ability of selected boiling water reactor (BWR) process line valves to perform their containment isolation function at high energy pipe break conditions and other more normal flow conditions. The fluid and valve operating responses were measured to provide information concerning valve and operator performance at various valve loadings so that the information could be used to assess typical nuclear industry motor operator sizing equations. Six valves were tested, three 6-in. isolation valves representative of those used in reactor water cleanup systems in BWRs and three 10-in. isolation valves representative of those used in BWR high pressure coolant injection (HPCI) steam lines. The concern with these normally open isolation valves is whether they will close in the event of a downstream pipe break outside of containment. The results of this testing will provide part of the technical insights for NRC efforts regarding Generic Issue 87 (GI-87), Failure of the HPCI Steam Line Without Isolation, which includes concerns about the uncertainties in gate valve motor operator sizing and torque switch settings for these BWR containment isolation valves. As of this writing, the Phase 2 test program has just been completed. Preliminary observations made in the field confirmed most of the results from the Phase 1 test program. All six valves closing in high energy water, high energy steam, and high pressure cold water require more force to close than would be calculated using the typical variables in the standard industry motor operator sizing equations
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
Mays, G.T.; Harrington, K.H.
206 forced shutdowns and power reductions were reviewed, along with 631 reportable events and other miscellaneous documentation concerning the operation of Dresden-2, in order to indicate those areas of plant operation that compromised plant safety. The most serious plant challenge to plant safety occurred on June 5, 1970; while undergoing power testing at 75% power, a spurious signal in the reactor pressure control system caused a turbine trip followed by a reactor scram. Subsequent erratic water level and pressure control in the reactor vessel, compounded by a stuck indicator pen on a water level monitor-recorder and inability of the isolation condenser to function, led to discharge of steam and water through safety valves into the reactor drywell. No significant contamination was discharged. There was no pressure damage or the reactor vessel of the drywell containment walls. Six areas of operation that should be of continued concern are diesel generator failures, control rod and rod drive malfunctions, radioactive waste management/health physics program problems, operator errors, turbine control valve and EHC problems, and HPCI failures. All six event types have continued to recur
DeWall, K.G.; Steele, R. Jr.
This report presents the results of research performed to develop technical insights for the NRC effort regarding Generic Issue 87, ''Failure of HPCI Steam Line Without Isolation.'' Volume III of this report contains the data and findings from the original research performed to assess the qualification of the valves and reported in EGG-SSRE-7387, ''Qualification of Valve Assemblies in High Energy BWR Systems Penetrating Containment.'' We present the original work here to complete the documentation trail. The recommendations contained in Volume III of this report resulted in the test program described in Volume I and II. The research began with a survey to characterize the population of normally open containment isolation valves in those process lines that connect to the primary system and penetrate containment. The qualification methodology used by the various manufacturers identified in the survey is reviewed and deficiencies in that methodology are identified. Recommendations for expanding the qualification of valve assemblies for high energy pipe break conditions are presented. 11 refs., 1 fig., 2 tabs
Gotoh, Toshiyuki; Watanabe, Takeshi
Recent experiments and Direct Numerical Simulations (DNSs) suggest that the small scale statistics of passive scalar may not be as ``universal'' as in the velocity case. To address this problem, we study the moments of scalar increment in steady turbulence at Rλ > 800 by using DNS up to the grid points of 40963. In order for the scalar and turbulent flow to be as faithful as possible to the assumptions that would be made in theories, Scalar 1 and Scalar 2 are simultaneously convected by the identical isotropic turbulent flow but excited by two different methods. Scalar 1 is excited by the random scalar injection which is isotropic, Gaussian and white in time at low wavenumber band, while Scalar 2 is excited by the uniform mean scalar gradient. The moments of two scalars as functions of the separation vector are expanded in terms of the Legendre polynomials to extract the scaling exponents of the moments up to the 4th anisotropic sector for Scalar 2. It is found that the exponents of the isotropic sectors seem to have the same values at separation distances in the narrow range over which the 4/3 law holds simultaneously for two scalars. The exponents of the anisotropic sectors and the cumulants of the moments will also be reported. HPCI, JHPCN, Grant-in-Aid for Sci. Res. No.24360068, Ministry of Edu. Sci., Japan.
Gotoh, Toshiyuki; Watanabe, Takeshi
It has long been considered that the moments of the scalar increment with separation distance r obey power law with scaling exponents in the inertial convective range and the exponents are insensitive to variation of pumping of scalar fluctuations at large scales, thus the scaling exponents are universal. We examine the scaling behavior of the moments of increments of passive scalars 1 and 2 by using DNS up to the grid points of 40963. They are simultaneously convected by the same isotropic steady turbulence atRλ = 805 , but excited by two different methods. Scalar 1 is excited by the random scalar injection which is isotropic, Gaussian and white in time at law wavenumber band, while Scalar 2 is excited by the uniform mean scalar gradient. It is found that the local scaling exponents of the scalar 1 has a logarithmic correction, meaning that the moments of the scalar 1 do not obey simple power law. On the other hand, the moments of the scalar 2 is found to obey the well developed power law with exponents consistent with those in the literature. Physical reasons for the difference are explored. Grants-in-Aid for Scientific Research 15H02218 and 26420106, NIFS14KNSS050, HPCI project hp150088 and hp140024, JHPCN project jh150012.
This paper reports that a motor operated valve (MOV) rebuild program at Peach Bottom Atomic power station began in October, 1986 with what is known internally as Modification (MOD) 1915. The Engineering the Research Department developed this modification to address requirements in NRC Bulletin 85-03. The MOD consisted of As found/As left testing of MOVs in the HPCI (high pressure coolant injection) and RCIC (reactor core isolation cooling) systems; six minor motor operator enhancements to facilitate maintenance and testing, and to increase reliability, and installation of a data acquisition network to support differential pressure testing of a select number of valves in Unit 2. Twenty-four valves were involved. Modification plans incorporated the work into the outage that was scheduled for December, 1986 to February, 1987. The plans took into account other preventive and corrective MOV maintenance tasks to be performed by the Maintenance Department. In addition, modifications of control circuits to satisfy separation criteria for Appendix R had to be integrated into the schedule. To facilitate testing, adjustments to the standard test methods under the Permits and Blocking System were necessary. The normal method of testing a piece of equipment after maintenance was to clear or temporarily clear the permit (red tag) and have a plant operator operate the equipment for the test group. This method for setting up the testing an MOV was considered unacceptable because it could occupy a plant operator for an entire shaft or longer
Brown, E.J.; Ashe, F.S.
The survey approach was to analyze several events and identify trends or patterns. The primary data source was licensee event reports (LERs) and consisted of 444 total valve operator events with 193 motor operator events which served as the basis for this study. The investigation revealed that motor-operated events could be grouped in three major categories which are torque switches, limit switches, and motors. The major findings are: (1) Torque switches do not appear to be a dominant cause of valve assembly inoperability. The reported information suggests torque switch events are an indication of symptomatic change with time in valve operability characteristics rather than a root cause of valve inoperability. (2) Repetitive problems are occurring with valve operators. It may occur on the same valve, a valve in similar service in a similar system, or a valve in similar service in a redundant train of the same system. (3) The plant operating staff objective appears to be a mode of finding measures to return inoperable equipment to operational status rather than to determine root causes of inoperability. (4) Motor burnout of valve motor operators has occurred quite frequently in High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems of BWR units. (orig./GL)
Choi, Y.A.; Feltus, M.A.
Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates
The author participated in international experts' meeting held by IAEA on May 21, 2013 and presented the paper focusing on human and organizational aspects of the Fukushima nuclear accident. It clarified TEPCO's basic recognition: 'The cause of the accident should not be treated merely as a natural disaster due to an enormous tsunami being something difficult to anticipate and we believe it is necessary to seriously acknowledge the result that TEPCO failed to avoid an accident which might have been avoided if ample preparations had been made in advance with thorough use of human intellect' and then reconsidered the Fukushima nuclear accident: 'could we predict an enormous tsunami and take whatever countermeasures?' and 'could we respond to the accident better?' for the worldwide operators to avoid such an accident, which moved meeting's participants deeply. Presentation's contents followed 'Reassessment of the Fukushima Nuclear Accident and Nuclear Safety Reform Plan' published by TEPCO on March 29. This article described outline of the presentation. Though the only way to explore the possibility to save Unit 1 was that operators could bravely go up to the 4th floor of reactor building and open the isolation valves to start IC, it was given up without any clear communication among key decision makers for confirming the IC operational status. As for Unit 3, operators could not achieve thorough focus on ensuring core cooling such that proactive transfer from RCIC/HPCI to low pressure water injection was not challenged, mainly because of low trust on Diesel/Driven Fire Protection Pump (DDFP). During the design stage and afterward, ample consideration was not given to common cause failures originating in external events, which led to a severe situation where almost all the power supplies and safety system functions were lost. Continuous efforts to reduce risks were not ample, including the collection, analysis and utilization of information on safety enhancement
Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Suzuki, Yoshio; Sawa, Kazuhiro; Iigaki, Kazuhiko
a structural analysis's code, which concerns interaction among components. This research and development is partially supported by HPCI strategic program of MEXT, Ministry of Education, Culture, Sports, and Science and Technology in Japan. (author)