CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

International Nuclear Information System (INIS)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author)

CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

Energy Technology Data Exchange (ETDEWEB)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author).

The effect of low capacity injection systems on transient initiated loss of vessel water injection at Browns Ferry unit one

International Nuclear Information System (INIS)

Ott, L.T.

1983-01-01

Probabilistic risk assessment (PRA) analyses have indicated the transient initiated loss of vessel water injection (TQUV sequence) to be a dominant accident scenario for BWR plants. The PRA studies assumed the low capacity injection systems to be unimportant in severe accidents. The results of a Severe Accident Sequence Analysis (SASA) Program study have shown that these systems are capable of preventing or significantly delaying core damage in a TQUV sequence

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

1985-07-01

Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

Computer Aided Design of The Cooling System for Plastic Injection Molds

Directory of Open Access Journals (Sweden)

Hakan GÜRÜN

2009-02-01

Full Text Available The design of plastic injection molds and their cooling systems affect both the dimension, the shape, the quality of a plastic part and the cycle time of process and the cost of mold. In this study, the solid model design of a plastic injection mold and the design of cooling sysytem were possibly carried out without the designer interaction. Developed program permited the use of three types of the cooling system and the different cavity orientations and the multible plastic part placement into the mold cores. The program which was developed by using Visual LISP language and the VBA (Visual BASIC for Application modules, was applicated in the AutoCAD software domain. Trial studies were presented that the solid model design of plastic injection molds and the cooling systems increased the reliability, the flexibility and the speed of the design.

A reavaluation of the reliability analysis of the low pressure injection system for Angra-1

International Nuclear Information System (INIS)

Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.; Tayt-Sohn, L.C.

1983-01-01

The emergency core cooling system of Angra 1 is analysed aiming at the low pressure injection systems, using the fault tree technique. All the failure mode of the components are considered for this analyse. (author) [pt

SQL injection detection system

OpenAIRE

Vargonas, Vytautas

2017-01-01

SQL injection detection system Programmers do not always ensure security of developed systems. That is why it is important to look for solutions outside being reliant on developers. In this work SQL injection detection system is proposed. The system analyzes HTTP request parameters and detects intrusions. It is based on unsupervised machine learning. Trained by regular request data system detects outlier user parameters. Since training is not reliant on previous knowledge of SQL injections, t...

Study of risk reduction by improving operation of reactor core isolation cooling system

International Nuclear Information System (INIS)

Watanabe, Yamato; Tazai, Ayuko; Yamagishi, Shohei; Muramatsu, Ken; Muta, Hitoshi

2014-01-01

The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

Hydrogen injection device in BWR type reactor

International Nuclear Information System (INIS)

Takagi, Jun-ichi; Kubo, Koji.

1988-01-01

Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

Probabilistic risk assessment (PRA) on the effectiveness of a core rescue system (SSN) for PWRs

International Nuclear Information System (INIS)

Petrangeli, G.; Valeri, A.

1983-01-01

Safety systems for the prevention of LWR core severe damage have recently been studied, which are based on automatic primary system depressurization and on borated water injection by low pressure accumulators. These systems have been named Core Rescue System (SSN). The present study evaluates the reduction in core melt probability brought about by the installation of a SSN system on the RSS (WASH 1400) PWR plant (Surry 1). The calculated result is a core melt probability reduction factor of about 250. Taking into account the possible effect of external or internal unknown events of negligible, yet undefined, probability it is concluded that a SSN system can make a plant ten times safer. The first part of a review report by Prof. N.C.Rasmussen, MIT, dealing with general comment, is attached

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Successful Treatment of Early Talar Osteonecrosis by Core Decompression Combined with Intraosseous Stem Cell Injection: A Case Report.

Science.gov (United States)

Nevalainen, Mika T; Repo, Jussi P; Pesola, Maija; Nyrhinen, Jukka P

2018-01-01

Osteonecrosis of the talus is a fairly rare condition. Many predisposing factors have been identified including previous trauma, use of corticosteroids, alcoholism, and smoking. As a gold standard, magnetic resonance imaging (MRI) is the most sensitive and specific diagnostic examination to detect osteonecrosis. While many treatment options for talar osteonecrosis exist, core decompression is suggested on young patients with good outcome results. More recently, intraosseous stem cell and platelet-rich plasma (PRP) injection has been added to the core decompression procedure. We report a successful treatment of early talar osteonecrosis ARCO I (Association Research Circulation Osseous) by core decompression combined with stem cell and PRP injection. On 3-month and 15-month follow-up, MRI showed complete resolution of the osteonecrotic changes together with clinical improvement. This modified technique is a viable treatment option for early talar osteonecrosis. Nevertheless, future prospects should include a study comparing this combined technique with plain core decompression.

A fast alternative to core plug tests for optimising injection water salinity for EOR

DEFF Research Database (Denmark)

Hassenkam, Tue; Andersson, Martin Peter; Hilner, Emelie Kristin Margareta

2014-01-01

of the clays which would lead to permanent reservoir damage but evidence of effectiveness at moderate salinity would offer the opportunity to dispose of produced water. The goal is to define boundary conditions so injection water salinity is high enough to prevent reservoir damage and low enough to induce...... the low salinity effect while keeping costs and operational requirements at a minimum. Traditional core plug testing for optimising conditions has some limitations. Each test requires a fresh sample, core testing requires sophisticated and expensive equipment, and reliable core test data requires several...... experiments can be done relatively quickly on very little material, it gives the possibility of testing salinity response on samples from throughout a reservoir and for gathering statistics. Our approach provides a range of data that can be used to screen core plug testing conditions and to provide extra data...

THE RHIC INJECTION SYSTEM.

Energy Technology Data Exchange (ETDEWEB)

FISCHER,W.; GLENN,J.W.; MACKAY,W.W.; PTITSIN,V.; ROBINSON,T.G.; TSOUPAS,N.

1999-03-29

The RHIC injection system has to transport beam from the AGS-to-RHIC transfer line onto the closed orbits of the RHIC Blue and Yellow rings. This task can be divided into three problems. First, the beam has to be injected into either ring. Second, once injected the beam needs to be transported around the ring for one turn. Third, the orbit must be closed and coherent beam oscillations around the closed orbit should be minimized. We describe our solutions for these problems and report on system tests conducted during the RHIC Sextant test performed in 1997. The system will be fully commissioned in 1999.

Emergency core cooling systems

International Nuclear Information System (INIS)

Kubokoya, Takashi; Okataku, Yasukuni.

1984-01-01

Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

ITER Neutral Beam Injection System

International Nuclear Information System (INIS)

Ohara, Yoshihiro; Tanaka, Shigeru; Akiba, Masato

1991-03-01

A Japanese design proposal of the ITER Neutral Beam Injection System (NBS) which is consistent with the ITER common design requirements is described. The injection system is required to deliver a neutral deuterium beam of 75MW at 1.3MeV to the reactor plasma and utilized not only for plasma heating but also for current drive and current profile control. The injection system is composed of 9 modules, each of which is designed so as to inject a 1.3MeV, 10MW neutral beam. The most important point in the design is that the injection system is based on the utilization of a cesium-seeded volume negative ion source which can produce an intense negative ion beam with high current density at a low source operating pressure. The design value of the source is based on the experimental values achieved at JAERI. The utilization of the cesium-seeded volume source is essential to the design of an efficient and compact neutral beam injection system which satisfies the ITER common design requirements. The critical components to realize this design are the 1.3MeV, 17A electrostatic accelerator and the high voltage DC acceleration power supply, whose performances must be demonstrated prior to the construction of ITER NBI system. (author)

Laboratory Mid-frequency (Kilohertz) Range Seismic Property Measurements and X-ray CT Imaging of Fractured Sandstone Cores During Supercritical CO2 Injection

Science.gov (United States)

Nakagawa, S.; Kneafsey, T. J.; Chang, C.; Harper, E.

2014-12-01

During geological sequestration of CO2, fractures are expected to play a critical role in controlling the migration of the injected fluid in reservoir rock. To detect the invasion of supercritical (sc-) CO2 and to determine its saturation, velocity and attenuation of seismic waves can be monitored. When both fractures and matrix porosity connected to the fractures are present, wave-induced dynamic poroelastic interactions between these two different types of rock porosity—high-permeability, high-compliance fractures and low-permeability, low-compliance matrix porosity—result in complex velocity and attenuation changes of compressional waves as scCO2 invades the rock. We conducted core-scale laboratory scCO2 injection experiments on small (diameter 1.5 inches, length 3.5-4 inches), medium-porosity/permeability (porosity 15%, matrix permeability 35 md) sandstone cores. During the injection, the compressional and shear (torsion) wave velocities and attenuations of the entire core were determined using our Split Hopkinson Resonant Bar (short-core resonant bar) technique in the frequency range of 1-2 kHz, and the distribution and saturation of the scCO2 determined via X-ray CT imaging using a medical CT scanner. A series of tests were conducted on (1) intact rock cores, (2) a core containing a mated, core-parallel fracture, (3) a core containing a sheared core-parallel fracture, and (4) a core containing a sheared, core-normal fracture. For intact cores and a core containing a mated sheared fracture, injections of scCO2 into an initially water-saturated sample resulted in large and continuous decreases in the compressional velocity as well as temporary increases in the attenuation. For a sheared core-parallel fracture, large attenuation was also observed, but almost no changes in the velocity occurred. In contrast, a sample containing a core-normal fracture exhibited complex behavior of compressional wave attenuation: the attenuation peaked as the leading edge of

Design of shutdown system no.2 liquid poison injection system for 500 MWe PHWR

International Nuclear Information System (INIS)

Bhatnagar, S.; Balasubrahmanian, A.K.; Pillai, A.V.

1997-01-01

Defence in depth and two group system concepts form the basic design philosophy for the shutdown systems. There are two independent, diverse and fast acting shutdown systems provided for the 500 MWe PHWR. The design is based on fail-safe principle, sufficient component redundancy and on-line testing. Liquid poison injection system, as shutdown system 2, is newly developed for the 500 MWe PHWRs. The system operates by rapidly injecting gadolinium nitrate solution into bulk moderator using stored helium pressure thereby inserting negative reactivity. A high pressure helium supply tank which provides the energy for system actuation, is connected, through an array of fast acting valves in series-parallel arrangement, to the individual poison tanks storing gadolinium nitrate solution. The valves, belonging to three different channels of reactor Protection System 2, are the only active components in the system. The valves are fail safe and are periodically tested on-line without actually firing the system. The system comprising of in-core assemblies and the external process system has been engineered. Experimental work is being carried out by BARC for design validation and data generation. This paper describes the conceptual development, design basis, design parameters and detailed engineering of the system. (author)

Effect of capillary number on the oil recovery using oil-water emulsion injection in core flooding experiments

Energy Technology Data Exchange (ETDEWEB)

Guillen Nunez, Victor Raul; Carvalho, Marcio da Silveira [Pontifical Catholic University of Rio de Janeiro (PUC-Rio), RJ (Brazil). Dept. of Mechanical Engineering], E-mail: msn@puc-rio.br; Basante, Vladimir Alvarado [University of Wyoming, Laramie, WY (United States). Dept. of Chemical/Petroleum Engineering], E-mail: valvard@uwyo.edu

2010-07-01

The Water injection flooding is a common method to improve reservoir sweep and pressure maintenance. The heavy-oil-recovery efficiency is in part limited by the high water-to-oil mobility ratio. Several enhanced oil recovery methods are being developed as more efficient alternatives to water flooding. Dispersion injection, in particular oil-water emulsion injection, has been tried with relative success as an enhanced oil recovery method, but the technique is not fully developed or understood. If emulsion injection proves to be an effective EOR method, its use would bring the added benefit of disposing produced water with small oil content that could be modified to serve as the injected oil-water emulsion. The use of such methods requires a detailed analysis of the different flow regimes of emulsions through the porous space of a reservoir rock. If the drop size of the disperse phase is of the same order of magnitude as the pore size, the drops may agglomerate and partially block water flow through pores. This flow regime may be used to control the mobility of the injected liquid, leading to higher recovery factor. We have shown in recent experiments of oil displacement in a sandstone core that, the oil recovery factor could be raised from approximately 40 %, obtained with water injection only, up to approximately 75 % by alternating water and emulsion injection. Although these results clearly show the improvement in the recovery factor, the mechanisms responsible for the phenomenon have not been clearly elucidated. In this work, two sandstone cores were used to demonstrate the effect of flow rate (capillary number) on the mobility control by emulsion injection. Figure 1 shows a schematic representation of the experiment set-up. The experiments show that raising the flow rate by a factor of 10 (0.03 ml/min to 0.3 ml/min), the oil recovered factor decreases considerably. (author)

[Strategy of constructing post-market integral evaluation system of traditional Chinese medicine injection].

Science.gov (United States)

Zhang, Xiao-Yu; Wang, Yan-Ping; Lin, Li-Kai; Shang, Hong-Cai; Wang, Yong-Yan

2017-08-01

As an important representative of modern Chinese medicine, traditional Chinese medicine (TCM) injzection has become an indispensable part of the Chinese medicine industry. However, its development is now restricted by the bottleneck of insufficient core competitiveness, low-level research and production, even injection quality and the safe use are not guaranteed. Thus, it is urgent to reevaluate post-marketing TCM injection generally and to make secondary development. Under current circumstances, taking major brands which have good clinical and market foundation, as well as research value, as the main subject of cultivation and evaluation is an important approach to innovative development of TCM injection industry. Unlike oral proprietary Chinese medicine, the cultivatation of major brands of TCM injection needs higher technical support, quality standards and more timely feedback. Therefore, a post-market integral evaluation system adaptive to TCM injection is required. This article discussed some key points on the construction of a post-market integral evaluation system of TCM injection in three levels： optimizing evaluation methods, building synergistic innovation platforms which combine the medical research institutions and pharmaceutical enterprises, and finally constructing the integral evaluation system. A "five to one" structure has been proposed to enhance TCM injection effectiveness, safety and adaptability on the whole, which are from the following aspects： mechanism research, clinical evidence validation, literature information mining, sustainable development of resources and industrialization operation. Copyright© by the Chinese Pharmaceutical Association.

Analysis of large break loss of coolant accident with simultaneous injection into cold leg and hot leg

International Nuclear Information System (INIS)

Luo Bangqi

1997-01-01

When a large break loss of coolant accident occurs, the most part of the safety injection water injected into the cold leg by the safety injection system will flow through the channel between the pressure vessel and the barrel out of the break into the containment, only a little part of the safety injection water can flow into the reactor core. If the safety injection can inject into both the cold leg and the hot leg simultaneously, the safety injection water injected from the cold leg will flow into the core more easily, because the safety injection water injected from the hot leg will carry out more heat from the upper plenum and the core, so the upper plenum and the core is depressed. In addition, a small part of the safety injection water injected from the hot leg will flow down in the core after impinging the guide tubes in the upper plenum, so the core will get more safety injection water than only cold leg injection, and the core will be much safer

Modern Cored Wire Injection 2PE-9 Method in the Production of Ductile Iron

Directory of Open Access Journals (Sweden)

E. Guzik

2012-04-01

Full Text Available The results of studies on the use of modern two cored wires injection method for production of nodular graphite cast iron with use of unique implementation of drum ladle as a treatment/ transport and casting ladle instead vertical treatment ladle was described. The injection of length of Ø 9mm wires, cored: in FeSi + Mg nodulariser mixture and inoculant master alloy is a treatment method which can be used to produce iron melted in coreless induction furnace. This paper describes the results of using this method for possibility production of ductile iron under specific industrial conditions. In this case was taken ductile iron with material designation: EN-GJS-450- 10 Grade according PN-EN 1563:2000. Microstructure of 28 trials was controlled on internally used sample which has been correlated with standard sample before. The paper presents typical metallic matrix and graphite characteristic. Additionally, mechanical properties were checked in one experiment. Because of further possibility treatment temperature reduction only the rough magnesium recovery and cost of this new method are given.

Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

International Nuclear Information System (INIS)

Choi, Y.A.; Feltus, M.A.

1995-01-01

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

Performance of ARCHITECT HCV core antigen test with specimens from US plasma donors and injecting drug users.

Science.gov (United States)

Mixson-Hayden, Tonya; Dawson, George J; Teshale, Eyasu; Le, Thao; Cheng, Kevin; Drobeniuc, Jan; Ward, John; Kamili, Saleem

2015-05-01

Hepatitis C virus (HCV) core antigen is a serological marker of current HCV infection. The aim of this study was mainly to evaluate the performance characteristics of the ARCHITECT HCV core antigen assay with specimens from US plasma donors and injecting drug users. A total of 551 serum and plasma samples with known anti-HCV and HCV RNA status were tested for HCV core antigen using the Abbott ARCHITECT HCV core antigen test. HCV core antigen was detectable in 100% of US plasma donor samples collected during the pre-seroconversion phase of infection (anti-HCV negative/HCV RNA positive). Overall sensitivity of the HCV core antigen assay was 88.9-94.3% in samples collected after seroconversion. The correlation between HCV core antigen and HCV RNA titers was 0.959. HCV core antigen testing may be reliably used to identify current HCV infection. Published by Elsevier B.V.

Environmental mitigation for SCC initiation of BWR core internals by hydrogen injection during start-up

International Nuclear Information System (INIS)

Dozaki, K.; Abe, A.; Nagata, N.; Takiguchi, H.

2004-01-01

Hydrogen injection into the reactor water has been applied to many BWR power stations. Since hydrogen injected accelerates recombination of oxidant generated by water radiolysis, oxidant concentration, such as dissolved oxygen concentration in reactor water can be reduced. As the result of the reduction of oxidant concentration, Electrochemical Corrosion Potential (ECP) at the surface of structural material can be lowered. Lowered ECP moderates Stress Corrosion Cracking (SCC) sensitivity of structural materials, such as stainless steels. As usual, hydrogen injection system begins to work after the plant start-up is finished, when the condition of normal operation is established. Accordingly, Hydrogen Water Chemistry (HWC) does not cover all the period of plant operation. As far as SCC crack growth is considered, loss of HWC during plant start-up does not result in significant crack growth, because of duration of plant start-up is much shorter than that of plant normal operation, when HWC condition is being satisfied. However, the reactor water environment and load conditions during a plant start-up may contribute to the initiation of SCC. It is estimated that the core internals are subjected to the strain rate that may cause susceptibility to SCC initiation during start-up. Dissolved oxygen (DO) and hydrogen peroxide (H 2 O 2 ) has a peak, and ECP is in high levels during start-up. Therefore it is beneficial to perform hydrogen injection during start-up as well in order to suppress SCC initiation. We call it HWC During Start-up (HDS) here. (orig.)

Diversified emergency core cooling in CANDU with a passive moderator heat rejection system

Energy Technology Data Exchange (ETDEWEB)

Spinks, N [AECL Research, Chalk River Labs., Chalk River, ON (Canada)

1996-12-01

A passive moderator heat rejection system is being developed for CANDU reactors which, combined with a conventional emergency-coolant injection system, provides the diversity to reduce core-melt frequency to order 10{sup -7} per unit-year. This is similar to the approach used in the design of contemporary CANDU shutdown systems which leads to a frequency of order 10{sup -8} per unit-year for events leading to loss of shutdown. Testing of a full height 1/60 power-and-volume-scaled loop has demonstrated the feasibility of the passive system for removal of moderator heat during normal operation and during accidents. With the frequency of core-melt reduced, by these measures, to order 10{sup -7} per unit year, no need should exist for further mitigation. (author). 3 refs, 2 figs.

Injection system of compact SR light source 'AURORA'

International Nuclear Information System (INIS)

Takayama, Takeshi; Yano, Takashi; Sasaki, Yasushi; Yasumitsu, Naoki

1991-01-01

A half-integer-resonance injection method is introduced for a superconducting SR-ring of 1 m orbit diameter, which is made of a weak focussing single-body magnet. The present method makes it possible to inject an electron beam of an energy of as high as 150 MeV into the ring of a magnetic field strength of 1 T. Several new injection devices are introduced in order to guide the beam under the strong magnetic fringing field, and to excite the half-integer-resonance. The field index of 0.73 is selected for the half-integer-resonance injection. The field index of 0.35 at the maximum magnetic field strength of 4.3 T is to get a sufficiently long quantum lifetime. A new device named resonance jumper is used to pass quickly several resonances of betatron motion without beam loss. The resonances occur when the magnetic field is ramped up and the field index decreases from 0.73 to 0.35. The injection devices except the inflector are air-core magnets in order to work in the strong magnetic field. In November of 1989, the beam was successfully injected and stored. The injection devices and the half-integer-resonance injection method were established. (author)

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1978-05-01

Of the total 10 ROSA-II/UHI performance tests, 6 were reported previously. The rest are presented and discussion is made on the effects of heat generation in the core and UHI injection and repeatability of experiments. In addition, the following are described: (1) Pressure spikes observed in the upper head after sudden stoppage of UHI injection, and (2) discharge flow oscillation possibly due to UHI water injection into the upper plenum. (auth.)

Dual fuel injection piggyback controller system

Science.gov (United States)

Muji, Siti Zarina Mohd.; Hassanal, Muhammad Amirul Hafeez; Lee, Chua King; Fawzi, Mas; Zulkifli, Fathul Hakim

2017-09-01

Dual-fuel injection is an effort to reduce the dependency on diesel and gasoline fuel. Generally, there are two approaches to implement the dual-fuel injection in car system. The first approach is changing the whole injector of the car engine, the consequence is excessive high cost. Alternatively, it also can be achieved by manipulating the system's control signal especially the Electronic Control Unit (ECU) signal. Hence, the study focuses to develop a dual injection timing controller system that likely adopted to control injection time and quantity of compressed natural gas (CNG) and diesel fuel. In this system, Raspberry Pi 3 reacts as main controller unit to receive ECU signal, analyze it and then manipulate its duty cycle to be fed into the Electronic Driver Unit (EDU). The manipulation has changed the duty cycle to two pulses instead of single pulse. A particular pulse mainly used to control injection of diesel fuel and another pulse controls injection of Compressed Natural Gas (CNG). The test indicated promising results that the system can be implemented in the car as piggyback system. This article, which was originally published online on 14 September 2017, contained an error in the acknowledgment section. The corrected acknowledgment appears in the Corrigendum attached to the pdf.

Core cooling system for reactor

International Nuclear Information System (INIS)

Kondo, Ryoichi; Amada, Tatsuo.

1976-01-01

Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

Emergency core cooling system

International Nuclear Information System (INIS)

Sato, Akira; Kobayashi, Masahide.

1983-01-01

Purpose: To enable a stable operation of an emergency core cooling system by preventing the system from the automatic stopping at an abnormally high level of the reactor water during its operation. Constitution: A pump flow rate signal and a reactor water level signal are used and, when the reactor water level is increased to a predetermined level, the pump flow rate is controlled by the reactor water level signal instead of the flow rate signal. Specifically, when the reactor water level is gradually increased by the water injection from the pump and exceeds a setting signal for the water level, the water level deviation signal acts as a demand signal for the decrease in the flow rate of the pump and the output signal from the water level controller is also decreased depending on the control constant. At a certain point, the output signal from the water level controller becomes smaller than the output signal from the flow rate controller. Thus, the output signal from the water level controller is outputted as the output signal for the lower level preference device. In this way, the reactor water level and the pump flow rate can be controlled within a range not exceeding the predetermined pump flow rate. (Horiuchi, T.)

A volumetric flow sensor for automotive injection systems

International Nuclear Information System (INIS)

Schmid, U; Krötz, G; Schmitt-Landsiedel, D

2008-01-01

For further optimization of the automotive power train of diesel engines, advanced combustion processes require a highly flexible injection system, provided e.g. by the common rail (CR) injection technique. In the past, the feasibility to implement injection nozzle volumetric flow sensors based on the thermo-resistive measurement principle has been demonstrated up to injection pressures of 135 MPa (1350 bar). To evaluate the transient behaviour of the system-integrated flow sensors as well as an injection amount indicator used as a reference method, hydraulic simulations on the system level are performed for a CR injection system. Experimentally determined injection timings were found to be in good agreement with calculated values, especially for the novel sensing element which is directly implemented into the hydraulic system. For the first time pressure oscillations occurring after termination of the injection pulse, predicted theoretically, could be verified directly in the nozzle. In addition, the injected amount of fuel is monitored with the highest resolution ever reported in the literature

A volumetric flow sensor for automotive injection systems

Science.gov (United States)

Schmid, U.; Krötz, G.; Schmitt-Landsiedel, D.

2008-04-01

For further optimization of the automotive power train of diesel engines, advanced combustion processes require a highly flexible injection system, provided e.g. by the common rail (CR) injection technique. In the past, the feasibility to implement injection nozzle volumetric flow sensors based on the thermo-resistive measurement principle has been demonstrated up to injection pressures of 135 MPa (1350 bar). To evaluate the transient behaviour of the system-integrated flow sensors as well as an injection amount indicator used as a reference method, hydraulic simulations on the system level are performed for a CR injection system. Experimentally determined injection timings were found to be in good agreement with calculated values, especially for the novel sensing element which is directly implemented into the hydraulic system. For the first time pressure oscillations occurring after termination of the injection pulse, predicted theoretically, could be verified directly in the nozzle. In addition, the injected amount of fuel is monitored with the highest resolution ever reported in the literature.

The TFTR 40 MW neutral beam injection system and DT operations

International Nuclear Information System (INIS)

Stevenson, T.; O'Connor, T.; Garzotto, V.

1995-01-01

Since December 1993, TFTR has performed DT experiments using tritium fuel provided mainly by neutral beam injection. Significant alpha particle populations and reactor-like conditions have been achieved at the plasma core, and fusion output power has risen to a record 10.7 MW using a record 40 MW NB heating. Tritium neutral beams have injected into over 480 DT plasmas and greater than 500 kCi have been processed through the neutral beam gas, cryo, and vacuum systems. Beam tritium injections, as well as tritium feedstock delivery and disposal, have now become part of routine operations. Shot reliability with tritium is about 90% and is comparable to deuterium shot reliability. This paper describes the neutral beam DT experience including the preparations, modifications, and operating techniques that led to this high level of success, as well as the critical differences in beam operations encountered during DT operations. Also, the neutral beam maintenance and repair history during DT operations, the corrective actions taken, and procedures developed for handling tritium contaminated components are discussed in the context of supporting a continuous DT program

CONCEPTUAL DESIGN OF THE NSLS-II INJECTION SYSTEM.

Energy Technology Data Exchange (ETDEWEB)

SHAFTAN,T.; ROSE, T.; PINAYEV, I.; HEESE, R.; BENGTSSON, J.; SKARITKA, J.; MENG, W.; OZAKI, S.; MEIER, R.; STELMACH, C.; LITVINENKO, V.; PJEROV, S.; SHARMA, S.; GANETIS, G.; HSEUH, H.C.; JOHNSON, E.D.; TSOUPAS, N.; GUO, W.; BEEBE-WANG, J.; LUCCIO, A.U.; YU, L.H.; RAPARIA, D.; WANG, D.

2007-06-25

We present the conceptual design of the NSLS-II injection system [1,2]. The injection system consists of a low-energy linac, booster and transport lines. We review two different injection system configurations; a booster located in the storage ring tunnel and a booster housed in a separate building. We briefly discuss main parameters and layout of the injection system components.

Role of passive valves & devices in poison injection system of advanced heavy water reactor

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

2014-01-01

The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water (H 2 O) cooled, heavy water (D 2 O) moderated reactor. The reactor design is based on well-proven water reactor technologies and incorporates a number of passive safety features such as natural circulation core cooling; direct in-bundle injection of light water coolant during a Loss of Coolant Accident (LOCA) from Advanced Accumulators and Gravity Driven Water Pool by passive means; Passive Decay Heat Removal using Isolation Condensers, Passive Containment Cooling System and Passive Containment Isolation System. In addition to above, there is another passive safety system named as Passive Poison Injection System (PPIS) which is capable of shutting down the reactor for a prolonged time. It is an additional safety system in AHWR to fulfill the shutdown function in the event of failure of wired shutdown systems i.e. primary and secondary shut down systems of the reactor. When demanded, PPIS injects the liquid poison into the moderator by passive means using passive valves and devices. On increase of main heat transport (MHT) system pressure beyond a predetermined value, a set of rupture disks burst, which in-turn actuate the passive valve. The opening of passive valve initiates inrush of high pressure helium gas into poison tanks to push the poison into the moderator system, thereby shutting down the reactor. This paper primarily deals with design and development of Passive Poison Injection System (PPIS) and its passive valves & devices. Recently, a prototype DN 65 size Poison Injection Passive Valve (PIPV) has been developed for AHWR usage and tested rigorously under simulated conditions. The paper will highlight the role of passive valves & devices in PPIS of AHWR. The design concept and test results of passive valves along with rupture disk performance will also be covered. (author)

The Effect of Temperature and Injection Rate during Water Flooding Using Carbonate Core Samples: An Experimental Approach

Directory of Open Access Journals (Sweden)

Yaser Ahmadi

2016-10-01

Full Text Available In many reservoirs, after water flooding, a large volume of oil is still left behind. Hot water injection is the most basic type of thermal recovery which increase recovery by improved sweep efficiency and thermal expansion of crude.In the present work, the effects of injection rate and the temperature of the injected water were surveyed by using core flooding apparatus. Water flooding was performed at different rates (0.2, 0.3, and 0.4 cc/min and temperatures (20 and 90 °C, and the reservoir temperature was about 63 °C. Oil recovery during hot water injection was more than water injection. Moreover, it was concluded that at injection rates of 0.2, 0.3, and 0.4 cc/min breakthrough time in hot water injection occurred 10 min later in comparison to water injection. The results showed that higher oil recovery and longer breakthrough time were obtained as a result of reducing injection rate. In the first 50 minutes, the oil recovery at injection rates of 0.2, 0.3 and 0.4 cc/min was 27.5, 34, and 46% respectively. It was found that at the beginning of injection, thermal and non-thermal injection recovery factors are approximately equal. Moreover, according to the results, recovery factor at the lowest rate in hot water (T=90 °C and q=0.2 cc/min is the best condition to obtain the highest recovery.

Measurement of electrical impedance of a Berea sandstone core during the displacement of saturated brine by oil and CO2 injections

Science.gov (United States)

Liu, Yu; Xue, Ziqiu; Park, Hyuck; Kiyama, Tamotsu; Zhang, Yi; Nishizawa, Osamu; Chae, Kwang-seok

2015-12-01

Complex electrical impedance measurements were performed on a brine-saturated Berea sandstone core while oil and CO2 were injected at different pressures and temperatures. The saturations of brine, oil, and CO2 in the core were simultaneously estimated using an X-ray computed tomography scanner. The formation factor of this Berea core and the resistivity indexes versus the brine saturations were calculated using Archie's law. The experimental results found different flow patterns of oil under different pressures and temperatures. Fingers were observed for the first experiment at 10 MPa and 40 °C. The fingers were restrained as the viscosity ratio of oil and water changed in the second (10 MPa and 25 °C) and third (5 MPa and 25 °C) experiments. The resistivity index showed an exponential increase with a decrease in brine saturation. The saturation exponent varied from 1.4 to 4.0 at different pressure and temperature conditions. During the oil injection procedure, the electrical impedance increased with oil saturation and was significantly affected by different oil distributions; therefore, the impedance varied whether the finger was remarkable or not, even if the oil saturation remained constant. During the CO2 injection steps, the impedance showed almost no change with CO2 saturation because the brine in the pores became immobile after the oil injection.

Triggering Collapse of the Presolar Dense Cloud Core and Injecting Short-lived Radioisotopes with a Shock Wave. V. Nonisothermal Collapse Regime

Energy Technology Data Exchange (ETDEWEB)

Boss, Alan P., E-mail: aboss@carnegiescience.edu [Department of Terrestrial Magnetism, Carnegie Institution for Science, 5241 Broad Branch Road, NW, Washington, DC 20015-1305 (United States)

2017-08-01

Recent meteoritical analyses support an initial abundance of the short-lived radioisotope (SLRI) {sup 60}Fe that may be high enough to require nucleosynthesis in a core-collapse supernova, followed by rapid incorporation into primitive meteoritical components, rather than a scenario where such isotopes were inherited from a well-mixed region of a giant molecular cloud polluted by a variety of supernovae remnants and massive star winds. This paper continues to explore the former scenario, by calculating three-dimensional, adaptive mesh refinement, hydrodynamical code (FLASH 2.5) models of the self-gravitational, dynamical collapse of a molecular cloud core that has been struck by a thin shock front with a speed of 40 km s{sup −1}, leading to the injection of shock front matter into the collapsing cloud through the formation of Rayleigh–Taylor fingers at the shock–cloud intersection. These models extend the previous work into the nonisothermal collapse regime using a polytropic approximation to represent compressional heating in the optically thick protostar. The models show that the injection efficiencies of shock front materials are enhanced compared to previous models, which were not carried into the nonisothermal regime, and so did not reach such high densities. The new models, combined with the recent estimates of initial {sup 60}Fe abundances, imply that the supernova triggering and injection scenario remains a plausible explanation for the origin of the SLRIs involved in the formation of our solar system.

Injection and Dump Systems

CERN Document Server

Bracco, C; Barnes, M J; Carlier, E; Drosdal, L N; Goddard, B; Kain, V; Meddahi, M; Mertens, V; Uythoven, J

2012-01-01

Performance and failures of the LHC injection and ex- traction systems are presented. In particular, a comparison with the 2010 run, lessons learnt during operation with high intensity beams and foreseen upgrades are described. UFOs, vacuum and impedance problems related to the injection and extraction equipment are analysed together with possible improvements and solutions. New implemented features, diagnostics, critical issues of XPOC and IQC applications are addressed.

Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

International Nuclear Information System (INIS)

Chupin, A.; Hu, L. W.; Buongiorno, J.

2008-01-01

Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

Fiber-Based, Injection-Molded Optofluidic Systems: Improvements in Assembly and Applications

Directory of Open Access Journals (Sweden)

Marco Matteucci

2015-12-01

Full Text Available We present a method to fabricate polymer optofluidic systems by means of injection molding that allow the insertion of standard optical fibers. The chip fabrication and assembly methods produce large numbers of robust optofluidic systems that can be easily assembled and disposed of, yet allow precise optical alignment and improve delivery of optical power. Using a multi-level chip fabrication process, complex channel designs with extremely vertical sidewalls, and dimensions that range from few tens of nanometers to hundreds of microns can be obtained. The technology has been used to align optical fibers in a quick and precise manner, with a lateral alignment accuracy of 2.7 ± 1.8 μm. We report the production, assembly methods, and the characterization of the resulting injection-molded chips for Lab-on-Chip (LoC applications. We demonstrate the versatility of this technology by carrying out two types of experiments that benefit from the improved optical system: optical stretching of red blood cells (RBCs and Raman spectroscopy of a solution loaded into a hollow core fiber. The advantages offered by the presented technology are intended to encourage the use of LoC technology for commercialization and educational purposes.

Optimization of the injection molding process for development of high performance calcium oxide -based ceramic cores

Science.gov (United States)

Zhou, P. P.; Wu, G. Q.; Tao, Y.; Cheng, X.; Zhao, J. Q.; Nan, H.

2018-02-01

The binder composition used for ceramic injection molding plays a crucial role on the final properties of sintered ceramic and to avoid defects on green parts. In this study, the effects of binder compositions on the rheological, microstructures and the mechanical properties of CaO based ceramic cores were investigated. It was found that the optimized formulation for dispersant, solid loading was 1.5 wt% and 84 wt%, respectively. The microstructures, such as porosity, pore size distribution and grain boundary density were closely related to the plasticizer contents. The decrease of plasticizer contents can enhance the strength of the ceramic cores but with decreased shrinkage. Meanwhile, the creep resistance of ceramic cores was enhanced by decreasing of plasticizer contents. The flexural strength of the core was found to decrease with the increase of the porosity, the improvement of creep resistance is closely related to the decrease of porosity and grain boundary density.

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-09-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented

Overview of core designs and requirements/criteria for core restraint systems

International Nuclear Information System (INIS)

Sutherland, W.H.

1984-01-01

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System is presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented. (author)

Core Science Systems--Mission overview

Science.gov (United States)

Gallagher, Kevin T.

2012-01-01

The Core Science Systems Mission Area delivers nationally focused Earth systems and information science that provides fundamental research and data that underpins all Mission Areas of the USGS, the USGS Science Strategy, and Presidential, Secretarial, and societal priorities. —Kevin T. Gallagher, Associate Director, Core Science Systems

Improved core protection calculator system algorithm

International Nuclear Information System (INIS)

Yoon, Tae Young; Park, Young Ho; In, Wang Kee; Bae, Jong Sik; Baeg, Seung Yeob

2009-01-01

Core Protection Calculator System (CPCS) is a digitized core protection system which provides core protection functions based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels which adapted a two out of four trip logic. CPCS algorithm improvement for the newly designed core protection calculator system, RCOPS (Reactor COre Protection System), is described in this paper. New features include the improvement of DNBR algorithm for thermal margin, the addition of pre trip alarm generation for auxiliary trip function, VOPT (Variable Over Power Trip) prevention during RPCS (Reactor Power Cutback System) actuation and the improvement of CEA (Control Element Assembly) signal checking algorithm. To verify the improved CPCS algorithm, CPCS algorithm verification tests, 'Module Test' and 'Unit Test', would be performed on RCOPS single channel facility. It is expected that the improved CPCS algorithm will increase DNBR margin and enhance the plant availability by reducing unnecessary reactor trips

SMART core protection system design

International Nuclear Information System (INIS)

Lee, J. K.; Park, H. Y.; Koo, I. S.; Park, H. S.; Kim, J. S.; Son, C. H.

2003-01-01

SMART COre Protection System(SCOPS) is designed with real-tims Digital Signal Processor(DSP) board and Network Interface Card(NIC) board. SCOPS has a Control Rod POSition (CRPOS) software module while Core Protection Calculator System(CPCS) consists of Core Protection Calculators(CPCs) and Control Element Assembly(CEA) Calculators(CEACs) in the commercial nuclear plant. It's not necessary to have a independent cabinets for SCOPS because SCOPS is physically very small. Then SCOPS is designed to share the cabinets with Plant Protection System(PPS) of SMART. Therefor it's very easy to maintain the system because CRPOS module is used instead of the computer with operating system

The PEP injection system

International Nuclear Information System (INIS)

Brown, K.L.; Avery, R.T.; Peterson, J.M.

1988-01-01

A system to transport 10-to-15-GeV electron and positron beams from the Stanford Linear Accelerator and to inject them into the PEP storage ring under a wide variety of lattice configurations has been designed. Optically, the transport line consists of three 360/degree/ phase-shift sections of FODO lattice, with bending magnets interspersed in such a way as to provide achromaticity, convenience in energy and emittance definition, and independent tuning of the various optical parameters for matching into the ring. The last 360/degree/ of phase shift has 88 milliradians of bend in a vertical plane and deposits the beam at the injection septum via a Lambertson magnet. Injection is accomplished by launching the beam with several centimeters of radial betatron amplitude in a fast bump provided by a triad of pulsed kicker magnets. Radiation damping reduces the collective amplitude quickly enough to allow injection at a high repetition rate

Low-power wide-locking-range injection-locked frequency divider for OFDM UWB systems

Energy Technology Data Exchange (ETDEWEB)

Yin Jiangwei; Li Ning; Zheng Renliang; Li Wei; Ren Junyan, E-mail: lining@fudan.edu.c [State Key Laboratory of ASIC and System, Fudan University, Shanghai 201203 (China)

2009-05-01

This paper describes a divide-by-two injection-locked frequency divider (ILFD) for frequency synthesizers as used in multiband orthogonal frequency division multiplexing (OFDM) ultra-wideband (UWB) systems. By means of dual-injection technique and other conventional tuning techniques, such as DCCA and varactor tuning, the divider demonstrates a wide locking range while consuming much less power. The chip was fabricated in the Jazz 0.18 mum RF CMOS process. The measurement results show that the divider achieves a locking range of 4.85 GHz (6.23 to 11.08 GHz) at an input power of 8 dBm. The core circuit without the test buffer consumes only 3.7 mA from a 1.8 V power supply and has a die area of 0.38 x 0.28 mm{sup 2}. The wide locking range combined with low power consumption makes the ILFD suitable for its application in UWB systems.

Low-power wide-locking-range injection-locked frequency divider for OFDM UWB systems

International Nuclear Information System (INIS)

Yin Jiangwei; Li Ning; Zheng Renliang; Li Wei; Ren Junyan

2009-01-01

This paper describes a divide-by-two injection-locked frequency divider (ILFD) for frequency synthesizers as used in multiband orthogonal frequency division multiplexing (OFDM) ultra-wideband (UWB) systems. By means of dual-injection technique and other conventional tuning techniques, such as DCCA and varactor tuning, the divider demonstrates a wide locking range while consuming much less power. The chip was fabricated in the Jazz 0.18 μm RF CMOS process. The measurement results show that the divider achieves a locking range of 4.85 GHz (6.23 to 11.08 GHz) at an input power of 8 dBm. The core circuit without the test buffer consumes only 3.7 mA from a 1.8 V power supply and has a die area of 0.38 x 0.28 mm 2 . The wide locking range combined with low power consumption makes the ILFD suitable for its application in UWB systems.

Polytopol computing for multi-core and distributed systems

Science.gov (United States)

Spaanenburg, Henk; Spaanenburg, Lambert; Ranefors, Johan

2009-05-01

Multi-core computing provides new challenges to software engineering. The paper addresses such issues in the general setting of polytopol computing, that takes multi-core problems in such widely differing areas as ambient intelligence sensor networks and cloud computing into account. It argues that the essence lies in a suitable allocation of free moving tasks. Where hardware is ubiquitous and pervasive, the network is virtualized into a connection of software snippets judiciously injected to such hardware that a system function looks as one again. The concept of polytopol computing provides a further formalization in terms of the partitioning of labor between collector and sensor nodes. Collectors provide functions such as a knowledge integrator, awareness collector, situation displayer/reporter, communicator of clues and an inquiry-interface provider. Sensors provide functions such as anomaly detection (only communicating singularities, not continuous observation), they are generally powered or self-powered, amorphous (not on a grid) with generation-and-attrition, field re-programmable, and sensor plug-and-play-able. Together the collector and the sensor are part of the skeleton injector mechanism, added to every node, and give the network the ability to organize itself into some of many topologies. Finally we will discuss a number of applications and indicate how a multi-core architecture supports the security aspects of the skeleton injector.

H- charge exchange injection systems

International Nuclear Information System (INIS)

Ankenbrandt, C.; Curtis, C.; Hojvat, C.; Johnson, R.P.; Owen, C.; Schmidt, C.; Teng, L.; Webber, R.C.

1980-01-01

The techniques and components required for injection of protons into cyclic accelerators by means of H - charge exchange processes are reviewed, with emphasis on the experience at Fermilab. The advantages of the technique are described. The design and performance of the system of injection of H - ions into the Fermilab Booster are detailed. (Auth.)

Economics of water injected air screw compressor systems

Science.gov (United States)

Venu Madhav, K.; Kovačević, A.

2015-08-01

There is a growing need for compressed air free of entrained oil to be used in industry. In many cases it can be supplied by oil flooded screw compressors with multi stage filtration systems, or by oil free screw compressors. However, if water injected screw compressors can be made to operate reliably, they could be more efficient and therefore cheaper to operate. Unfortunately, to date, such machines have proved to be insufficiently reliable and not cost effective. This paper describes an investigation carried out to determine the current limitations of water injected screw compressor systems and how these could be overcome in the 15-315 kW power range and delivery pressures of 6-10 bar. Modern rotor profiles and approach to sealing and cooling allow reasonably inexpensive air end design. The prototype of the water injected screw compressor air system was built and tested for performance and reliability. The water injected compressor system was compared with the oil injected and oil free compressor systems of the equivalent size including the economic analysis based on the lifecycle costs. Based on the obtained results, it was concluded that water injected screw compressor systems could be designed to deliver clean air free of oil contamination with a better user value proposition than the oil injected or oil free screw compressor systems over the considered range of operations.

Performance of the ALS injection system

International Nuclear Information System (INIS)

Kim, C.H.

1993-05-01

The authors started commissioning the Advanced Light Source (ALS) storage ring on January 11, 1993. The stored beam reached 60 mA on March 24, 1993 and 407 mA on April 9, 1993. The fast pace of storage ring commissioning can be attributed partially to the robust injection system. In this paper they describe the operating characteristics of the ALS injection system

Injection system of the minicyclotron accelerator mass spectrometer

International Nuclear Information System (INIS)

Liu Yonghao; Li Deming; Chen Maobai; Lu Xiangshun

1999-01-01

The existing injection system of the SMCAMS (super-sensitive mini-cyclotron accelerator mass spectrometer) is described together with the discussion of its disadvantages exposed after having been operating for five years, which provides a basis for consideration of improvements to the injection system. An optimized injection system with an analytical magnet added prior to the minicyclotron has been proposed and calculated

SCORPIO - VVER core surveillance system

International Nuclear Information System (INIS)

Zalesky, K.; Svarny, J.; Novak, L.; Rosol, J.; Horanes, A.

1997-01-01

The Halden Project has developed the core surveillance system SCORPIO which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. So far the system has only been implemented on western PWRs but the basic concept is applicable to a wide range of reactor including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO was initiated in cooperation with the Nuclear Research Institute at Rez and industry partners in the Czech Republic. The first system will be installed at the Dukovany NPP. (author)

Operational considerations for the PSB H- Injection System

CERN Document Server

Weterings, W; Borburgh, J; Carli, C; Fowler, T; Goddard, B

2010-01-01

For the LINAC4 project the PS Booster (PSB) injection system will be upgraded. The 160 MeV Hbeam will be distributed to the 4 superimposed PSB synchrotron rings and horizontally injected by means of an H- charge-exchange system. Operational considerations for the injection system are presented, including expected beam losses from unwanted field stripping of H- and excited H0 and foil scattering, possible injection failure cases and expected stripping foil lifetimes. Loading assumptions for the internal beam dumps are discussed together with estimates of doses on various components.

Installation of JMTR core management system

International Nuclear Information System (INIS)

Imaizumi, Tomomi; Ide, Hiroshi; Naka, Michihiro; Komukai, Bunsaku; Nagao, Yoshiharu

2013-01-01

In order to carry out the core management after the reoperation of JMTR quickly and accurately, the authors took up the Standard Reactor Analysis Code (SRAC) system and core management support programs that are operating in a general-purpose large computer and transferred them to PC (OS: Linux), and newly established a JMTR core management system. As for the core analysis, this measure enabled an increase in the processing speed from the check of core arrangement to the result display of nuclear restriction values to about 60 times, compared with the conventional method. It was confirmed that the differences of calculation results originated from the difference of internal display of computers, associated with the transfer of each analysis code from GS21-400 system to PC-Linux, were within practically allowable level. In the future, this system will be applied to the core analysis of JMTR, as well as to the preparation of operation plans. (A.O.)

Injection system of teh SSC Medium Energy Booster

International Nuclear Information System (INIS)

Mao, N.; Gerig, R.; McGill, J.; Brown, K.

1994-04-01

The Medium Energy Booster (MEB) is the third of the SSCL accelerators and the largest of the resistive magnet synchrotrons. It accelerates protons from an injection momentum of 12 GeV/c to a top momentum of 200 GeV/c. A beam injection system has been designed to inject the beam transferred from the Low Energy Booster onto the MEB closed orbit in the MEB injection insertion region. The beam is injected via a vertical bending Lambertson septum magnet and a horizontal kicker with appropriate matching and very little beam loss and emittance dilution. The beam optics of the injection system is described in this paper. The required parameters of the Lambertson septum magnet and the injection kicker are given

Preparation of Fe(3)O(4)@C@CNC multifunctional magnetic core/shell nanoparticles and their application in a signal-type flow-injection photoluminescence immunosensor.

Science.gov (United States)

Chu, Chengchao; Li, Meng; Li, Long; Ge, Shenguang; Ge, Lei; Yu, Jinghua; Yan, Mei; Song, Xianrang

2013-11-01

We describe here the preparation of carbon-coated Fe3O4 magnetic nanoparticles that were further fabricated into multifunctional core/shell nanoparticles (Fe3O4@C@CNCs) through a layer-by-layer self-assembly process of carbon nanocrystals (CNCs). The nanoparticles were applied in a photoluminescence (PL) immunosensor to detect the carcinoembryonic antigen (CEA), and CEA primary antibody was immobilized onto the surface of the nanoparticles. In addition, CEA secondary antibody and glucose oxidase were covalently bonded to silica nanoparticles. After stepwise immunoreactions, the immunoreagent was injected into the PL cell using a flow-injection PL system. When glucose was injected, hydrogen peroxide was obtained because of glucose oxidase catalysis and quenched the PL of the Fe3O4@C@CNC nanoparticles. The here proposed PL immunosensor allowed us to determine CEA concentrations in the 0.005–50 ng·mL-1 concentration range, with a detection limit of 1.8 pg·mL-1.

SCORPIO - WWER core surveillance system

International Nuclear Information System (INIS)

Hornaes, Arne; Bodal, Terje; Sunde, Svein; Zalesky, K.; Lehman, M.; Pecka, M.; Svarny, J.; Krysl, V.; Juzova, Z.; Sedlak, A.; Semmler, M.

1998-01-01

The Institut for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(Authors)

SCORPIO - VVER core surveillance system

International Nuclear Information System (INIS)

Hornaes, A.; Bodal, T.; Sunde, S.

1998-01-01

The Institutt for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators, which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including VVERs. The main differences between VVERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a VVER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(author)

Impact of physical properties of biodiesel on the injection process in a common-rail direct injection system

International Nuclear Information System (INIS)

Boudy, Frederic; Seers, Patrice

2009-01-01

This paper presents the influence of biodiesel fuel properties on the injection mass flow rate of a diesel common-rail injection system. Simulations are first performed with ISO 4113 diesel fuel on a four-cylinder common-rail system to evaluate a single and triple injection strategies. For each injection strategy, the impact of modifying a single fuel property at a time is evaluated so as to quantify its influence on the injection process. The results show that fuel density is the main property that affects the injection process, such as total mass injected and pressure wave in the common-rail system. The fuel's viscosity and bulk modulus also influence, but to a lessen degree, the mass flow rate of the injector notably during multiple injection strategies as individual properties change the fuel's dampening property and friction coefficient.

Optimum design of the injection duct system of a stenter machine

Energy Technology Data Exchange (ETDEWEB)

Juraeva, Makhsuda; Song, Dong Joo [Yeungnam University, Geyongsan (Korea, Republic of); Ryu, Kyung Jin [Ajou Motor College, Boryeong (Korea, Republic of)

2017-05-15

Stenter machines are used for drying fabrics in the textile industry and have a heater, injection duct system, and fans inside a chamber. The injection duct system has ducts and air-injecting holes. Plane-type injection duct systems were investigated to obtain uniform airflow at the air-injecting holes. The flow field of the injection duct systems was computed using ANSYS CFX with different heights of the duct end and different shapes for the air-injecting holes. There was a high mass flow rate at the air-injecting holes and high airflow circulation inside both plane-type and mountain-type ducts at the ends. The height of the duct end was varied between 40 mm and 160 mm. The injection duct systems were analyzed with four different shapes of air-injecting holes. The circular and elliptical holes had lower standard deviations of the mass flow rate than other shapes. Relatively uniform mass flow rates were obtained in the plane-type and mountain-type duct systems when the height of the duct end was 40 mm and the shape of the air-injecting holes was circular or elliptical. The developed injection duct systems were improved by obtaining a uniform mass flow rate at the air-injecting holes. A stenter prototype was fabricated with the developed injection duct system to confirm the numerical results. The developed injection duct system had better performance than the original system.

Development of CANDU core monitoring system

International Nuclear Information System (INIS)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H.

2003-01-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1

Development of CANDU core monitoring system

Energy Technology Data Exchange (ETDEWEB)

Yoon, M. Y.; Yeam, C. S.; Kwon, O. H.; Kim, K. H. [Institute for Advanced Engineering, Yongin (Korea, Republic of)

2003-07-01

The research was performed to develop a CANDU Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong Nuclear Power Plant(NPP) No. 1. CCMS uses RFSP(Reactor Fueling Simulation Program) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from DCC(Digital Control Computer) for the purpose of producing basic input data. CCMS could be largely divided into two modules; CCMS server program and CCMS client program. CCMS server program plays the role in automatic and continuous RFSP run and management of the past output data resulting from the run using Data Base Management System(DBMS). CCMS client program enables users to monitor current and past core status with GUI(Graphic-User Interface) environment predefined. The effectiveness of CCMS was verified by comparing the data resulted from field-test of the system for about 43 hours with the data used in the field of Wolsong NPP No. 1.

Validation of reactor core protection system

International Nuclear Information System (INIS)

Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

2008-01-01

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

Assessment of core thermo-hydrodynamic models of REFLA-1D with CCTF data

International Nuclear Information System (INIS)

Okubo, Tsutomu; Murao, Yoshio

1983-07-01

In order to assess the core thermo-hydrodynamic models of REFLA-1D/MODE3, which is the latest version of REFLA-1D, several calculations of the core thermo-hydrodynamics have been performed for the CCTF Core-I series tests. The measured initial and boundary conditions were used for these calculations. The calculational results showed that the water accumulation model of Case 2 could predict the CCTF results fairly well as it could for the JAERI small scale facility. The calculated results for the base case and the EM tests were in good agreement with the CCTF data. The parameter effects, such as system pressure, initial clad temperature, Acc injection rate, LPCI injection rate and initial down-comer wall temperature, were predicted correctly, except for the high system pressure and the high LPCI injection rate tests. (author)

Performance Analysis of Multi Stage Safety Injection Tank

International Nuclear Information System (INIS)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo

2015-01-01

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Performance Analysis of Multi Stage Safety Injection Tank

Energy Technology Data Exchange (ETDEWEB)

Shin, Soo Jai; Kim, Young In; Bae, Youngmin; Kang, Han-Ok; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In general the integral reactor has such characteristics, the integral reactor requires a high flow rate of coolant safety injection at the initial stage of the accident in which the core level is relatively fast decreased, A medium flow rate of coolant safety injection at the early and middle stages of the accident in which the coolant discharge flow rate is relatively large due to a high internal pressure of the reactor vessel, and a low flow rate of coolant safety injection is required at the middle and late stages of the accident in which the coolant discharge flow rate is greatly reduced due to a decreased pressure of the reactor vessel. It is noted that a high flow rate of the integral reactor is quite smaller compared to a flow rate required in the commercial loop type reactor. However, a nitrogen pressurized safety injection tank has been typically designed to quickly inject a high flow rate of coolant when the internal pressure of the reactor vessel is rapidly decreased, and a core makeup tank has been designed to safely inject at a single mode flow rate due to a gravitational head of water subsequent to making a pressure balance between the reactor vessel and core makeup tank. As a result, in order to compensate such a disadvantage, various type systems are used in a complicated manner in a reactor according to the required characteristic of safety injection during an accident. In the present study, we have investigated numerically the performance of the multi stage safety injection tank. A parameter study has performed to understand the characteristics of the multi stage safety injection tank. The performance of the multi stage safety injection tank has been investigated numerically. When an accident occurs, the coolant in the multi stage safety injection tank is injected into a reactor vessel by a gravitational head of water subsequent to making a pressure balance between the reactor and tank. At the early stages of the accident, the high flow rate of

Facile Phosphine-Free Synthesis of CdSe/ZnS Core/Shell Nanocrystals Without Precursor Injection

Directory of Open Access Journals (Sweden)

Zhu Chang-Qing

2008-01-01

Full Text Available AbstractA new simple method for synthesis of core/shell CdSe/ZnS nanocrystals (NCs is present. By adapting the use of cadmium stearate, oleylamine, and paraffin liquid to a non-injection synthesis and by applying a subsequent ZnS shelling procedure to CdSe NCs cores using Zinc acetate dihydrate and sulfur powder, luminescent CdSe/ZnS NCs with quantum yields of up to 36% (FWHM 42–43 nm were obtained. A seeding-growth technique was first applied to the controlled synthesis of ZnS shell. This method has several attractive features, such as the usage of low-cost, green, and environmentally friendlier reagents and elimination of the need for air-sensitive, toxic, and expensive phosphines solvent. Furthermore, due to one-pot synthetic route for CdSe/ZnS NCs, the approach presented herein is accessible to a mass production of these NCs.

Improving safety margin of LWRs by rethinking the emergency core cooling system criteria and safety system capacity

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2016-10-15

Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.

Water injection device for reactor container

International Nuclear Information System (INIS)

Sakaki, Isao.

1996-01-01

A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

Perspectives on Severe Accident Management by Depressurization and External Water Injection under Extended SBO Conditions

International Nuclear Information System (INIS)

Seol, Wookcheol; Park, Jongwoon

2014-01-01

Three major issues of severe accident management guideline (SAMG) after this sort of extended SBO would be depressurization of the primary system, external water injection and hydrogen management inside a containment. Under this situation, typical SAM actions would be depressurization and external water delivery into the core. However, limited amount of external water would necessitate optimization between core cooling, containment integrity and fission product removal. In this paper, effects of SAM actions such as depressurization and external water injection on the reactor and containment conditions after extended SBO are analyzed using MAAP4 code. Positive and negative aspects are discussed with respect to core cooling and fission product retention inside a primary system. Conclusions are made as following: Firstly, early depressurization action itself has two-faces: positive with respect to delay of the reactor vessel failure but negative with respect to the containment failure and fission product retention inside the primary system. Secondly, in order to prevent containment overpressure failure after external water injection, re-closing of PORV later should be considered in SAM, which has never been considered in the previous SAMG. Finally, in case of external water injection, the flow rate should be optimized considering not only the cooling effect but also the long term fission product retention inside the primary system

Influence green sand system by core sand additions

Directory of Open Access Journals (Sweden)

N. Špirutová

2012-01-01

Full Text Available Today, about two thirds of iron alloys casting (especially for graphitizing alloys of iron are produced into green sand systems with usually organically bonded cores. Separation of core sands from the green sand mixture is very difficult, after pouring. The core sand concentration increase due to circulation of green sand mixture in a closed circulation system. Furthermore in some foundries, core sands have been adding to green sand systems as a replacement for new sands. The goal of this contribution is: “How the green sand systems are influenced by core sands?”This effect is considered by determination of selected technological properties and degree of green sand system re-bonding. From the studies, which have been published yet, there is not consistent opinion on influence of core sand dilution on green sand system properties. In order to simulation of the effect of core sands on the technological properties of green sands, there were applied the most common used technologies of cores production, which are based on bonding with phenolic resin. Core sand concentration added to green sand system, was up to 50 %. Influence of core sand dilution on basic properties of green sand systems was determined by evaluation of basic industrial properties: moisture, green compression strength and splitting strength, wet tensile strength, mixture stability against staling and physical-chemistry properties (pH, conductivity, and loss of ignition. Ratio of active betonite by Methylene blue test was also determined.

A new dual injection system for AMS facility

International Nuclear Information System (INIS)

Liu Lin; Zhou Weijian; Cheng Peng; Yu Huagui; Chen Maobai

2007-01-01

In order to measure long-lived radioisotopes such as 10 Be with high sensitivity using an HVEE model 4130 AMS system, as well as to guarantee 14 C measurements of high precision, a new dual injection system for the AMS system is proposed. The proposal is to add a Wien filter located between the ion source system and the recombinator of the HVEE model 4130. When a pulsing voltage is optionally applied to the Wien filter, a sequential injection mode is turned on. The isotopes would alternately pass on different trajectories through the recombinator. When the pulsing voltage and magnetic field are turned off, the Wien filter acts as a field-free drift space and the standard simultaneous injection mode is on. Beam optics calculation show that the new dual injection system will increase the number of radio-nuclides which can be analyzed, keep the high precision capability for radiocarbon dating and achieve high sensitivity for 10 Be and 26 Al measurements, together with simplifying the layout as compared to existing dual-injector and dual high-energy beam line systems

Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.N.; Mohan, L.R.

2002-01-01

Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

Dimethyl Ether in Diesel Fuel Injection Systems

DEFF Research Database (Denmark)

Sorenson, Spencer C.; Glensvig, M.; Abata, D. L.

1998-01-01

A study of the behaviour of DME in diesel injection systems. A discussion of the effects of compressibility of DME on compression work and wave propagation.DME spray shapes and penetration rates......A study of the behaviour of DME in diesel injection systems. A discussion of the effects of compressibility of DME on compression work and wave propagation.DME spray shapes and penetration rates...

Safety assessment for the ultimate heat sink (UHS) system with non-injection concept in nuclear power plants (NPPs)

International Nuclear Information System (INIS)

Kim, Yun Il; Woo, Tae Ho

2017-01-01

Following the Fukushima accident, it is proposed to find a better safety system, which has a pool-type cooling system without coolant injections. Since the conventional piping-based injection systems have failed in treating the three major severe accidents, the artificial pool could be constructed to cover the failed reactor core systems in which the pool-like structure is constructed. Regarding this study, there were some previous studies about the ultimate heat sink (UHS). In this study, the system dynamics (SD) modeling is performed in the case of Fukushima Unit 1 accident. The basic events are obtained by the Boolean values as 0 and 1. The quantifications are obtained by the SD algorithm incorporated with the Vensim software. In the simulations work, there is a plateau region between the 25th and 45th years in the interested period. The nonlinear algorithm is applied for the UHS analysis which was not installed for the commercial use yet. (author)

Los Alamos Proton Storage Ring (PSR) injection deflector system

International Nuclear Information System (INIS)

Jason, A.j.; Higgins, E.F.; Koelle, A.R.

1983-01-01

We describe a pulsed magnetic deflector system planned for the injection system of the PSR. Two sets of magnets, appropriately placed in the optical systems of both the ring and the injection transport line, provide control of the rate at which particles are injected into a given portion of transverse phase space and limit the interaction of stored beam with the injection stripping foil. High-current modulators that produce relatively complex waveforms are required for this purpose. Solid-state drivers using direct feedback to produce the necessary waveforms are discussed as replacements for the more conventional high-voltage tube technology

PLT neutral beam injection systems

International Nuclear Information System (INIS)

Menon, M.M.; Barber, G.C.; Blue, C.W.

1979-01-01

A brief description of the Princeton Large Torus (PLT) neutral beam injection system is given and its performance characteristics are outlined. A detailed operational procedure is included, as are some tips on troubleshooting. Proper operation of the source is shown to be a crucial factor in system performance

Functional requirements for core surveillance systems

International Nuclear Information System (INIS)

Andersson, T.

2000-01-01

Operating experience at Ringhals-2 has demonstrated the feasibility of a mixed core surveillance system comprised of fixed in-core detectors combined with the original movable detector system. A small number of fixed in-core detectors provide continuous measurement of the thermal margins while the movable detectors are used mainly at start-up to verify the expected power distribution. Reactor noise diagnostics and neural networks can further improve the monitoring system. The reliability of the movable detector system can be improved by mechanical simplification. Wear and maintenance costs are lowered if the required flux-mapping frequency is reduced. Improved computer codes make the measurement uncertainties less dependent on the number of instrumented positions. A mixed system requires new types of technical specifications. (author)

Gaseous poison injection device

International Nuclear Information System (INIS)

Kubota, Ryuji; Sugisaki, Toshihiko; Inada, Ikuo.

1983-01-01

Purpose: To rapidly control the chain reaction due to thermal neutrons in a reactor core by using gaseous poisons as back-up means for control rod drives. Constitution: Gaseous poisons having a large neutron absorption cross section are used as back-up means for control rod drives. Upon failure of control rod insertion, the gaseous poisons are injected into the lower portion of the reactor core to control the reactor power. As the gaseous poisons, vapors at a high temperature and a higher pressure than that of the coolants in the reactor core are injected to control the reactor power due to the void effects. Since the gaseous poisons thus employed rapidly reach the reactor core and form gas bubbles therein, the deccelerating effect of the thermal neutrons is decreased to reduce the chain reaction. (Moriyama, K.)

Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-03-01

Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

Core sampling system spare parts assessment

International Nuclear Information System (INIS)

Walter, E.J.

1995-01-01

Soon, there will be 4 independent core sampling systems obtaining samples from the underground tanks. It is desirable that these systems be available for sampling during the next 2 years. This assessment was prepared to evaluate the adequacy of the spare parts identified for the core sampling system and to provide recommendations that may remediate overages or inadequacies of spare parts

Mini-cavity plasma core reactors for dual-mode space nuclear power/propulsion systems

International Nuclear Information System (INIS)

Chow, S.

1976-01-01

A mini-cavity plasma core reactor is investigated for potential use in a dual-mode space power and propulsion system. In the propulsive mode, hydrogen propellant is injected radially inward through the reactor solid regions and into the cavity. The propellant is heated by both solid driver fuel elements surrounding the cavity and uranium plasma before it is exhausted out the nozzle. The propellant only removes a fraction of the driver power, the remainder is transferred by a coolant fluid to a power conversion system, which incorporates a radiator for heat rejection. In the power generation mode, the plasma and propellant flows are shut off, and the driver elements supply thermal power to the power conversion system, which generates electricity for primary electric propulsion purposes

The Westinghouse BEACON on-line core monitoring system

International Nuclear Information System (INIS)

Buechel, Robert J.; Boyd, William A.; Casadei, Alberto L.

1995-01-01

BEACON (Best Estimate Analysis of Core Operations - Nuclear), a core monitoring and operational support package developed by Westinghouse, has been installed at many operating PWRs worldwide. The BEACON system is a real-time monitoring system which can be used in plants with both fixed and movable incore detector systems and utilizes an on-line nodal model combined with core instrumentation data to provide continuous core power distribution monitoring. In addition, accurate core-predictive capabilities utilizing a full core nodal model updated according to plant operating history can be made to provide operational support. Core history information is kept and displayed to help operators anticipate core behavior and take pro-active control actions. The BEACON system has been licensed by the U.S. Nuclear Regulatory Commission for direct, continuous monitoring of DNBR and peak linear heat rate. This allows BEACON to be integrated into the plant technical specifications to permit significant relaxation of operating limitations defined by conventional technical specifications. (author). 4 refs, 2 figs, 1 tab

Overview of on-line core monitoring system BEACON

International Nuclear Information System (INIS)

Dai Qing; Chen Xiaosong

2013-01-01

After more than 20 years of development, key technologies embedded with such system have reached a certain degree of maturity among some foreign countries. However, domestically, there is no comparable system yet. Through in-depth research and analysis on the most widely used core monitoring system in the world, BEACON, it's hope that this will provide guidance on our independent development of the first core monitoring system in China. Excore detectors, core outlet thermocouples and incore movable detectors are used to provide measure data on the status of the core for BEACON. Under the assumption of nodal homogeneity, an effective fast group model is used to solve the diffusion equation, followed by core-wise interpolation by Green's function. Finally, reconstruction of a calculated core is fitted with measured data using the surface spline function. The most significant technological advances are core monitoring during unstable core conditions, the use of nodal expansion method to improve accuracy and the adoption of single point calibration to increase the period of recalibration for the whole core. (authors)

Microcontroller-driven fluid-injection system for atomic force microscopy.

Science.gov (United States)

Kasas, S; Alonso, L; Jacquet, P; Adamcik, J; Haeberli, C; Dietler, G

2010-01-01

We present a programmable microcontroller-driven injection system for the exchange of imaging medium during atomic force microscopy. Using this low-noise system, high-resolution imaging can be performed during this process of injection without disturbance. This latter circumstance was exemplified by the online imaging of conformational changes in DNA molecules during the injection of anticancer drug into the fluid chamber.

Piezoelectric Injection Systems

Science.gov (United States)

Mock, R.; Lubitz, K.

The origin of direct injection can be doubtlessly attributed to Rudolf Diesel who used air assisted injection for fuel atomisation in his first self-ignition engine. Although it became apparent already at that time that direct injection leads to reduced specific fuel consumption compared to other methods of fuel injection, it was not used in passenger cars for the moment because of its disadvantageous noise generation as the requirements with regard to comfort were seen as more important than a reduced specific consumption.

A PIV Study of Slotted Air Injection for Jet Noise Reduction

Science.gov (United States)

Henderson, Brenda S.; Wernet, Mark P.

2012-01-01

Results from acoustic and Particle Image Velocimetry (PIV) measurements are presented for single and dual-stream jets with fluidic injection on the core stream. The fluidic injection nozzles delivered air to the jet through slots on the interior of the nozzle at the nozzle trailing edge. The investigations include subsonic and supersonic jet conditions. Reductions in broadband shock noise and low frequency mixing noise were obtained with the introduction of fluidic injection on single stream jets. Fluidic injection was found to eliminate shock cells, increase jet mixing, and reduce turbulent kinetic energy levels near the end of the potential core. For dual-stream subsonic jets, the introduction of fluidic injection reduced low frequency noise in the peak jet noise direction and enhanced jet mixing. For dual-stream jets with supersonic fan streams and subsonic core streams, the introduction of fluidic injection in the core stream impacted the jet shock cell structure but had little effect on mixing between the core and fan streams.

Beacon-Colss core monitoring system application and benefits

International Nuclear Information System (INIS)

Boyd, W.A.; Yoon, T.Y.

2005-01-01

Westinghouse and KNFC are creating an upgraded core monitoring system by merging the BEACON system (best estimate analyzer for core operation-nuclear) and COLSS (core operating limit supervisory system) into an integrated product. Although both BEACON and COLSS are core monitoring systems that have been in operation at many plants for a number of years, they each have some features and capabilities that are not in the other. Therefore it has been decided to incorporate portions of COLSS into the beacon system to create an optional level to support core monitoring applications on selected combustion engineering (C-E) designed plants. This optional level in the beacon system will be called BEACON-COLSS and will allow the beacon system to monitor the LCO's and Tech Spec limits at CE plants that currently use COLSS. This paper will present the structure of the new core monitoring system and the benefits it achieves for current COLSS plants, i.e., CE plants in the US and KSNP (Korean standard nuclear power plant). (authors)

Vibration analysis of the Golfech 2 safety injection system

International Nuclear Information System (INIS)

Morilhat, P.

1993-01-01

The main function of the safety injection system in a PWR plant is to ensure cooling of fuel elements in the event of a loss of coolant accident. The multistage centrifugal pump mounted-on this system induces pressure fluctuations, resulting in dynamic loads on piping. In certain plant units, these loads have caused cracking in the nozzles connected to the safety injection system, whereas in others, no damage has been observed. In order to understand the differences in dynamic behavior observed from one site to another, tests were performed on a real safety injection system, that of Golfech-2. They enabled determination of the modal characteristics of the system and identification of the hydro-acoustic source of the low head safety injection pump. They also enabled assessment of the pressure fluctuation levels in the pump suction and discharge areas as well as the vibratory response of the system when operating under partial and nominal flow conditions. Finally, these test results were used to estimate fatigue damage in the safety injection system. The experimental results will later be used to validate the model of the system undertaken with the piping design code CIRCUS and define the boundary conditions to be taken into account. (author). 6 figs., 2 refs

Metabolic clearance rate and urinary clearance of purified beta-core

International Nuclear Information System (INIS)

Wehmann, R.E.; Blithe, D.L.; Flack, M.R.; Nisula, B.C.

1989-01-01

We injected a highly purified preparation of the beta-core molecule, a fragment of hCG beta excreted in pregnancy urine, into five men and three women to determine its kinetic parameters, MCR, and urinary clearance. The beta-core molecule was distributed in an initial volume [1950 +/- 156 (mean +/- SEM) mL/m2 body surface area] approximately equal to the estimated plasma volume. Its disappearance was multiexponential on a semilogarithmic plot, with a rapid phase t1/2 of 3.5 +/- 0.7 min and a slow phase t1/2 of 22.4 +/- 4.2 min. The transit time (the mean time spent by a molecule of beta-core in transit) was 20.6 +/- 2.1 min. The MCR was 192.0 +/- 8.0 mL/min.m2 body surface area. About 5% of the injected dose of beta-core was excreted into the urine in the first 30 min after injection, and low levels of excretion persisted for up to 7 days. The urinary clearance rate of beta-core was 13.7 +/- 1.4 mL/min.m2, accounting for about 8% of the elimination of beta-core from the plasma. The beta-core immunoreactivity in serum and urine was characterized by gel filtration and three independent RIA systems to show that its properties were indistinguishable from those of the injected beta-core. Serum levels of beta-core in pregnant women were less than 0.2 ng/mL, while the amounts excreted in their urine were as much as 5 mg/day. Based on these clearance parameters of beta-core in normal subjects, less than 0.2% of the beta-core excreted in pregnancy urine is derived by urinary clearance of plasma beta-core. Therefore, more than 99% of the beta-core excreted in pregnancy urine is derived from beta-core in a compartment separate from plasma. In particular, these data indicate that there is relatively little placental secretion of beta-core into plasma and that placental secretion does not account for the vast majority of beta-core in pregnancy urine

Post-LOCA core flushing system

International Nuclear Information System (INIS)

Boyajian, J.D.; Weinberger, P.A.

1980-01-01

A system is disclosed for flushing the core of a nuclear reactor after a loss-of-coolant accident. A pump causes flow of liquid-phase fluid from the containment-vessel sump. This flow is used to provide the motivating force for an eductor that causes suction at the hot log of the reactor. The eductor suction can draw gas-phase coolant out of the hot leg. As a result, it can reduce pressure which may be preventing the flow of liquid-phase coolant out of the hot leg. By causing liquid-phase flow through the reactor, the system ensures that particles and boric acid are flushed out of the core. The system thereby eliminates the build-up of particles and the concentrations of boric acid in the core that could result if the coolant were to leave the pressure vessel exclusively in the gas phase. 9 claims

NMR-MPar: A Fault-Tolerance Approach for Multi-Core and Many-Core Processors

Directory of Open Access Journals (Sweden)

Vanessa Vargas

2018-03-01

Full Text Available Multi-core and many-core processors are a promising solution to achieve high performance by maintaining a lower power consumption. However, the degree of miniaturization makes them more sensitive to soft-errors. To improve the system reliability, this work proposes a fault-tolerance approach based on redundancy and partitioning principles called N-Modular Redundancy and M-Partitions (NMR-MPar. By combining both principles, this approach allows multi-/many-core processors to perform critical functions in mixed-criticality systems. Benefiting from the capabilities of these devices, NMR-MPar creates different partitions that perform independent functions. For critical functions, it is proposed that N partitions with the same configuration participate of an N-modular redundancy system. In order to validate the approach, a case study is implemented on the KALRAY Multi-Purpose Processing Array (MPPA-256 many-core processor running two parallel benchmark applications. The traveling salesman problem and matrix multiplication applications were selected to test different device’s resources. The effectiveness of NMR-MPar is assessed by software-implemented fault-injection. For evaluation purposes, it is considered that the system is intended to be used in avionics. Results show the improvement of the application reliability by two orders of magnitude when implementing NMR-MPar on the system. Finally, this work opens the possibility to use massive parallelism for dependable applications in embedded systems.

Web-based Core Design System Development

International Nuclear Information System (INIS)

Moon, So Young; Kim, Hyung Jin; Yang, Sung Tae; Hong, Sun Kwan

2011-01-01

The selection of a loading pattern is one of core design processes in the operation of a nuclear power plant. A potential new loading pattern is identified by selecting fuels that to not exceed the major limiting factors of the design and that satisfy the core design conditions for employing fuel data from the existing loading pattern of the current operating cycle. The selection of a loading pattern is also related to the cycle plan of an operating nuclear power plant and must meet safety and economic requirements. In selecting an appropriate loading pattern, all aspects, such as input creation, code runs and result processes are processed as text forms manually by a designer, all of which may be subject to human error, such as syntax or running errors. Time-consuming results analysis and decision-making processes are the most significant inefficiencies to avoid. A web-based nuclear plant core design system was developed here to remedy the shortcomings of an existing core design system. The proposed system adopts the general methodology of OPR1000 (Korea Standard Nuclear Power Plants) and Westinghouse-type plants. Additionally, it offers a GUI (Graphic User Interface)-based core design environment with a user-friendly interface for operators. It reduces human errors related to design model creation, computation, final reload core model selection, final output confirmation, and result data validation and verification. Most significantly, it reduces the core design time by more than 75% compared to its predecessor

Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Feldman, E.M.

1977-03-01

Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

DIAGNOSTICS OF GASOLINE FUEL SYSTEMS WITH DIRECT INJECTION

Directory of Open Access Journals (Sweden)

M. Bulgakov

2017-11-01

Full Text Available A method of diagnosing fuel systems with direct injection by means of producing a pressure oscillation in a hydraulic accumulator is presented. Having obtained a signal from pressure sensor it is possible to register a pressure drop at the moment of injection. If the system has a malfunction, then the pressure drop will be higher.

Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Collins, B.L.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

The kicker magnet system for TRISTAN Accumulation Ring injection

International Nuclear Information System (INIS)

Sakamoto, Y.; Satoh, K.; Nakayama, H.

1994-12-01

The injection of electron beams to TRISTAN Accumulation Ring (AR) was started in November 1983 and the positron injection started in November 1985. For the injection of electron and positron beams to AR, the unique kicker system was developed. In the kicker power supply the charging to the main capacitor was done with the resonant charge system together with the auxiliary charging unit. The impedance matching circuit was added to the kicker magnet for getting the required current form with least reflecting oscillation. In this paper we report the performance of this kicker system. (author)

Fast Flux Test Facility core system

International Nuclear Information System (INIS)

Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

1990-11-01

A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

Evaluation report on CCTF CORE-I REFLOOD TEST Cl-4 (Run 13) and Cl-15 (Run 24)

International Nuclear Information System (INIS)

Sudoh, Takashi; Murao, Yoshio.

1983-08-01

The tests Cl-4 and Cl-15 were performed with the Cylindrical Core Test Facility (CCTF) to investigate the effects of the depressurization process to simulate the refill phase, and the effects of the nitrogen to be injected after the end of the accumulator injection on the thermo-hydraulic behavior in the core and primary loop system during refill and reflood phases. In these tests, after the lower plenum was filled to 0.9m level with saturated water at 0.6 MPa, the accumulator water was injected into three intact cold legs in the depressurization period from 0.6 MPa to 0.2 MPa. The water in the lower plenum voided during the depressurization and the significant steam condensation occurred in or near the intact cold legs. The condensation caused high steam flow rate in the intact loops and the lower plenum flashing resulted in suppressed core water accumulation. The slightly lower core heat transfer coefficient due to the less core water caused the higher turnaround temperature and the longer quench time than those of the normal reflood test without the depressurization process. The nitrogen injection followed the accumulator injection was allowed in the test Cl-15. However, significant effects of the nitrogen injection was not observed. (author)

Characteristics of pressure wave in common rail fuel injection system of high-speed direct injection diesel engines

Directory of Open Access Journals (Sweden)

Mohammad Reza Herfatmanesh

2016-05-01

Full Text Available The latest generation of high-pressure common rail equipment now provides diesel engines possibility to apply as many as eight separate injection pulses within the engine cycle for reducing emissions and for smoothing combustion. With these complicated injection arrangements, optimizations of operating parameters for various driving conditions are considerably difficult, particularly when integrating fuel injection parameters with other operating parameters such as exhaust gas recirculation rate and boost pressure together for evaluating calibration results. Understanding the detailed effects of fuel injection parameters upon combustion characteristics and emission formation is therefore particularly critical. In this article, the results and discussion of experimental investigations on a high-speed direct injection light-duty diesel engine test bed are presented for evaluating and analyzing the effects of main adjustable parameters of the fuel injection system on all regulated emission gases and torque performance. Main injection timing, rail pressure, pilot amount, and particularly pilot timing have been examined. The results show that optimization of each of those adjustable parameters is beneficial for emission reduction and torque improvement under different operating conditions. By exploring the variation in the interval between the pilot injection and the main injection, it is found that the pressure wave in the common rail has a significant influence on the subsequent injection. This suggests that special attentions must be paid for adjusting pilot timing or any injection interval when multi-injection is used. With analyzing the fuel amount oscillation of the subsequent injections to pilot separation, it demonstrates that the frequency of regular oscillations of the actual fuel amount or the injection pulse width with the variation in pilot separation is always the same for a specified fuel injection system, regardless of engine speed

Designing Fault-Injection Experiments for the Reliability of Embedded Systems

Science.gov (United States)

White, Allan L.

2012-01-01

This paper considers the long-standing problem of conducting fault-injections experiments to establish the ultra-reliability of embedded systems. There have been extensive efforts in fault injection, and this paper offers a partial summary of the efforts, but these previous efforts have focused on realism and efficiency. Fault injections have been used to examine diagnostics and to test algorithms, but the literature does not contain any framework that says how to conduct fault-injection experiments to establish ultra-reliability. A solution to this problem integrates field-data, arguments-from-design, and fault-injection into a seamless whole. The solution in this paper is to derive a model reduction theorem for a class of semi-Markov models suitable for describing ultra-reliable embedded systems. The derivation shows that a tight upper bound on the probability of system failure can be obtained using only the means of system-recovery times, thus reducing the experimental effort to estimating a reasonable number of easily-observed parameters. The paper includes an example of a system subject to both permanent and transient faults. There is a discussion of integrating fault-injection with field-data and arguments-from-design.

The Advanced Photon Source injection timing system

International Nuclear Information System (INIS)

Lenkszus, F.R.; Laird, R.

1995-01-01

The Advanced Photon Source consists of five accelerators. The injection timing system provides the signals required to cause a bunch emitted from the electron gun to navigate through intermediate accelerators to a specific bucket (1 out of 1296) within the storage ring. Two linacs and a positron accumulator ring operate at 60Hz while a booster synchrotron ramps and injects into the storage ring at 2Hz. The distributed, modular VME/VXI-based injection timing system is controlled by two EPICS-based input/output controllers (IOCs). Over 40 VME/VXI cards have been developed to implement the system. Card types range from 352MHz VXI timing modules to VME-based fiber optic fanouts and logic translators/drivers. All timing is distributed with fiber optics. Timing references are derived directly from machine low-level rf of 9.77MHz and 352MHz. The timing references provide triggers to programmable delay generators. Three grades of timing are provided. Precision timing is derived from commercial digital delay generators, intermediate precision timing is obtained from VXI 8-channel digital delay generators which provide timing with 25ns peak-to-peak jitter, and modest precision timing is provided by the APS event system. The timing system is fully integrated into the APS EPICS-based control system

Core reset system design for linear induction accelerator

International Nuclear Information System (INIS)

Durga Praveen Kumar, D.; Mitra, S.; Sharma, Archana; Nagesh, K.V.; Chakravarthy, D.P.

2006-01-01

A repetitive pulsed power system based Linear Induction Accelerator (LIA-200) is being developed at BARC to get an electron beam of 200keV, 5kA, 50ns, 10-100 Hz. Amorphous core is the heart of these accelerators. It serves various functions in different subsystems viz. pulse power modulator, pulse transformer, magnetic switches and induction cavities. One of the factors that make the magnetic components compact is utilization of the total flux swing available in the core. In the present system, magnetic switches, pulse transformers, and induction cavity are designed to avail the full flux swing available in the core. For achieving this objective, flux density in the core has to be kept at the reverse saturation, before the main pulse is applied. The electrical circuit which makes it possible is called the core reset system. In this paper the details of core reset system designed for LIA-200 are described. (author)

User interface design and system integration aspects of core monitoring systems

International Nuclear Information System (INIS)

Berg, O.; Bodal, T.; Hornaes, A.; Porsmyr, J.

2000-01-01

The present paper describes our experience with the SCORPIO core monitoring system using generic building blocks for the MMI and system integration. In this context the different layers of the software system are discussed starting with the communication system, interfacing of various modules (e.g. physics codes), administration of several modules and generation of graphical user interfaces for different categories of end-users. A method by which re-use of software components can make the system development and maintenance more efficient is described. Examples are given from different system installation projects. The methodology adopted is considered particularly important in the future, as it is anticipated that core monitoring systems will be expanded with new functions (e.g. information from technical specifications, procedures, noise analysis, etc). Further, efficient coupling of off-line tools for core physics calculations and on-line modules in core monitoring can pave the way for cost savings. (authors)

Diagnosing the PEP-II Injection System

Energy Technology Data Exchange (ETDEWEB)

Decker, F.-J.; Donald, M.H.; Iverson, R.H.; Kulikov, A.; Pappas, G.C.; Weaver, M.; /SLAC

2005-05-09

The injection of beam into the PEP-II B-Factory, especially into the High Energy Ring (HER) has some challenges. A high background level in the BaBar detector has for a while inhibited us from trickling charge into the HER similar to the Low Energy Ring (LER). Analyzing the injection system has revealed many issues which could be improved. The injection bump between two kickers was not closed, mainly because the phase advance wasn't exactly 180{sup o} and the two kicker strengths were not balanced. Additionally we found reflections which kick the stored beam after the main kick and cause the average luminosity to drop about 3% for a 10 Hz injection rate. The strength of the overall kick is nearly twice as high as the design, indicating a much bigger effective septum thickness. Compared with single beam the background is worse when the HER beam is colliding with the LER beam. This hints that the beam-beam force and the observed vertical blow-up in the HER pushes the beam and especially the injected beam further out to the edge of the dynamic aperture or beyond.

Diagnosing the PEP-II Injection System

International Nuclear Information System (INIS)

Decker, F.-J.; Donald, M.H.; Iverson, R.H.; Kulikov, A.; Pappas, G.C.; Weaver, M.; SLAC

2005-01-01

The injection of beam into the PEP-II B-Factory, especially into the High Energy Ring (HER) has some challenges. A high background level in the BaBar detector has for a while inhibited us from trickling charge into the HER similar to the Low Energy Ring (LER). Analyzing the injection system has revealed many issues which could be improved. The injection bump between two kickers was not closed, mainly because the phase advance wasn't exactly 180 o and the two kicker strengths were not balanced. Additionally we found reflections which kick the stored beam after the main kick and cause the average luminosity to drop about 3% for a 10 Hz injection rate. The strength of the overall kick is nearly twice as high as the design, indicating a much bigger effective septum thickness. Compared with single beam the background is worse when the HER beam is colliding with the LER beam. This hints that the beam-beam force and the observed vertical blow-up in the HER pushes the beam and especially the injected beam further out to the edge of the dynamic aperture or beyond

Foam injection method and system

Energy Technology Data Exchange (ETDEWEB)

Hardy, W C; Parmley, J B; Shepard, J C

1977-05-10

A method is described for more efficiently practicing in situ combustion techniques by generating a gas-water mist or foam adjacent to the combustion formation within the injection well. The mist or foam is forced out of the well into the formation to transport heat away from the burned region of the formation toward the periphery of the combustion region to conserve fuel. Also taught are a method and system for fluid treating a formation while maintaining enhanced conformance of the fluid injection profile by generating a mist or foam down-hole adjacent to the formation and then forcing the mist or foam out into the formation. (19 claims)

Prediction of Golden Time for Recovering the Safety Injection System in Severe LOCA Circumstances

International Nuclear Information System (INIS)

Yoo, Kwae Hwan; Kim, Dong Young; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun

2015-01-01

In this study, the core uncovery and RV failure according to LOCA break sizes were analyzed by using the MAAP4 code when safety injection system (SIS) was not operating normally. We predicted the golden time of SIS recovery for accomplishing the reactor cold shutdown and preventing RV failure. MAAP4 code was used for severe accident analysis. The LOCA simulations were performed with break size in order to predict the golden time to recovery SIS. We predicted the golden time according to the SIS operation cases through the simulation of OPR1000. When LOCA occurred, the normal operation of SIS is very important in maintaining the integrity of NPPs. However if the SIS does not work or its actuation is delayed due to failure of the equipment, the DBA will lead to a severe accident. In this study, accident situations that SIS does not work normally were assumed and a number of MAAP4 code simulations were conducted. In addition, core uncovery time and RV failure time were predicted. If the recovery time of SIS for accident recovery is predicted, the core will not be exposed through appropriate action

Welding lines formation in holes obtained by low pressure injection molding of ceramic parts

Directory of Open Access Journals (Sweden)

C. A. Costa

Full Text Available Abstract This work presents a study to evaluate the process of producing internal holes in ceramic disks produced by low pressure injection molding (LPIM process. Two process conditions defined as pre-injection and post-injection were used to test the proposition. In the first one the pin cores that produce the holes were positioned in the cavity before the injection of the feedstock; and in the second one, the pin cores were positioned in the cavity, just after the feeding phase of the injection mold. An experimental injection mold designed and manufactured to test both processes was developed to produce ceramic disk with Ø 50 x 2 mm with four holes of Ø 5 mm, equally and radially distributed through the disk. The feedstock was composed of 86 wt% alumina (Al2O3 and 14 wt% organic vehicle based on paraffin wax. Heating and cooling systems controlled by a data acquisition system were included in the mold. The results showed that there were no welding lines with the post-injection process, proving to be an option for creating holes in the ceramic parts produced by LPIM. It was observed that best results were obtained at 58 °C mold temperature. The pins extraction temperature was about 45 °C, and the injection pressure was 170 kPa.

OPTICALLY BASED CHARGE INJECTION SYSTEM FOR IONIZATION DETECTORS

International Nuclear Information System (INIS)

CHEN, H.; CITTERIO, M.; LANNI, F.; LEITE, M.A.L.; RADEKA, V.; RESCIA, S.; TAKAI, H.

2001-01-01

An optically coupled charge injection system for ionization based radiation detectors which allows a test charge to be injected without the creation of ground loops has been developed. An ionization like signal from an external source is brought into the detector through an optical fiber and injected into the electrodes by means of a photodiode. As an application example, crosstalk measurements on a liquid Argon electromagnetic calorimeter readout electrodes were performed

Evaluation of BEACON-COLSS Core Monitoring System Benefits

International Nuclear Information System (INIS)

Kim, Joon Sung; Park, Young Ho; Morita, Toshio; Book, Michael A.

2005-01-01

In Korean Standard Nuclear Power Plant COLSS (Core Operating Limit Supervisory System) is used to monitor the DNBR Power Operating Limit (DNBRPOL) and Linear Heat Rate POL (KWPFPOL). Westinghouse and KNFC have developed an upgraded core monitoring system by combining the BEACON TM core monitoring system 1 (Best Estimate Analyzer for Core Operation . Nuclear) and COLSS into an integrated product that is called BEACON-COLSS. BEACON-COLSS generates the 3-D power distribution corrected by the in-core detectors measurements. The 3-D core power distribution methodology in BEACON-COLSS is significantly better than the synthesis methodology in COLSS. BEACONCOLSS uses the CETOP-D 2 thermal hydraulic code instead of CETOP-1. CETOP-D is a multi-channel thermal hydraulics code that will provide more accurate DNBR calculations than the DNBR calculators currently used in COLSS

Multi-core fiber undersea transmission systems

DEFF Research Database (Denmark)

Nooruzzaman, Md; Morioka, Toshio

2017-01-01

Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components.......Various potential architectures of branching units for multi-core fiber undersea transmission systems are presented. It is also investigated how different architectures of branching unit influence the number of fibers and those of inline components....

Variable volume combustor with pre-nozzle fuel injection system

Science.gov (United States)

Keener, Christopher Paul; Johnson, Thomas Edward; McConnaughhay, Johnie Franklin; Ostebee, Heath Michael

2016-09-06

The present application provides a combustor for use with a gas turbine engine. The combustor may include a number of fuel nozzles, a pre-nozzle fuel injection system supporting the fuel nozzles, and a linear actuator to maneuver the fuel nozzles and the pre-nozzle fuel injection system.

Pressure Fluctuations in a Common-Rail Fuel Injection System

Science.gov (United States)

Rothrock, A M

1931-01-01

This report presents the results of an investigation to determine experimentally the instantaneous pressures at the discharge orifice of a common-rail fuel injection system in which the timing valve and cut-off valve were at some distance from the automatic fuel injection valve, and also to determine the methods by which the pressure fluctuations could be controlled. The results show that pressure wave phenomena occur between the high-pressure reservoir and the discharge orifice, but that these pressure waves can be controlled so as to be advantageous to the injection of the fuel. The results also give data applicable to the design of such an injection system for a high-speed compression-ignition engine.

Automation of Aditya vacuum control system based on CODAC Core System

Energy Technology Data Exchange (ETDEWEB)

Raulji, Vismaysinh D., E-mail: vismay@ipr.res.in; Pujara, Harshad; Arambhadiya, Bharat; Jadeja, Kumarpalsinh; Bhatt, Shailesh; Rajpal, Rachana

2016-11-15

Highlights: • Monitor and control of vacuum control system based on CODAC Core System. • Communication between SIEMENS PLC and open source software EPICS. • With CODAC Core easy to configure and programming of slow controller. - Abstract: The main objective of vacuum control system is to provide ultrahigh vacuum for Aditya Tokamak operations. Aditya Vacuum vessel is having four vacuum pumping lines. To demonstrate implementation of automation; a study case is under taken by automating single Pumping Line of the Aditya vacuum system using CODAC Core System (CCS). Currently, vacuum system is operated manually. The CCS based control system allows remote control, monitoring, alarm handling of vacuum parameters. The CODAC Core System is the Linux based software package that is distributed by ITER Organization for the development of Plant System I&C software. CODAC Core System includes EPICS, CSS (Control System Studio) etc. CSS is used for HMI (Human Machine Interface), alarms and archives. SDD (Self Description Data) tool is used to configure plant system I&C. SDD Editor is an Eclipse based application to define the plant system, interface, I&C component, interfaced signals, configure variable. SCADA (Supervisory Control and Data Acquisition) system is developed in CSS. Data is transferred between PLC and CSS through EPICS. The complete system is tested with Aditya Vacuum Control System with process interlocks. Operator interface is also developed using Lab VIEW as a choice of the user. This paper will describe the salient features of the developed control system in detail.

Design of a microscopic electrical impedance tomography system using two current injections

International Nuclear Information System (INIS)

Liu, Qin; Oh, Tong In; Wi, Hun; Woo, Eung Je; Lee, Eun Jung; Seo, Jin Keun

2011-01-01

We describe a novel design of a microscopic electrical impedance tomography (micro-EIT) system for long-term noninvasive monitoring of cell or tissue cultures. The core of the micro-EIT system is a sample container including two pairs of current-injection electrodes and 360 voltage-sensing electrodes. In designing the container, we took advantage of a hexagonal structure with fixed dimensions and electrode configuration. This eliminated technical difficulties related to the unknown irregular boundary geometry of an imaging object in conventional medical EIT. Attaching a pair of large current-injection electrodes fully covering the left and right sides of the hexagonal container, we generated uniform parallel current density inside the container filled with saline. The 360 voltage-sensing electrodes were placed on the front, bottom and back sides of the hexagonal container in three sets of 8 × 15 arrays with equal gaps between them. We measured voltage differences between all neighboring pairs along the direction of the parallel current pathway. For the homogeneous container, all measured voltages must be the same since the voltage changes linearly along that direction. Any anomaly in the container perturbed the current pathways and therefore equipotential lines to produce different differential voltage data. For conductivity image reconstructions, we adopted a lately developed image reconstruction algorithm for this electrode configuration to first produce projected conductivity images on the front, bottom and back sides. Using a backprojection method, we reconstructed three-dimensional conductivity images from those projection images. To improve the image quality and also to meet the mathematical requirement on the uniqueness of a reconstructed image, we used a second pair of thin and long current-injection electrodes located at the middle of the front and back sides. This paper describes the design and construction of such a micro-EIT system with experimental

ROSA/LSTF experiment report for RUN SB-CL-24 repeated core heatup phenomena during 0.5% cold leg break LOCA

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Mitsuhiro; Anoda, Yoshinari [Department of Reactor Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

2000-03-01

A small break loss-of-coolant accident (SBLOCA) in a Westinghouse-type four-loop PWR was simulated in an experiment (SB-CL-24) conducted at the Large-Scale Test Facility (LSTF) with an intention to study repeated core heatup during a long-term cooldown process. The experiment was conducted on February 28, 1990 with specified test conditions including failure assumptions both on the high pressure injection (HPI) and the auxiliary feedwater systems, and the intentional secondary system depressurization as an operator action. The secondary depressurization contributed to promote the primary depressurization and the actuation of accumulator injection system (AIS). A temporary core heatup was observed in each of three loopseal clearing (LSC) processes. A significant core heatup occurred in the following boil-off process after loss of the secondary coolant mass and the AIS termination due to increase of the primary pressure. By additional opening of the pressurizer relief valves and safety valves, the primary pressure rapidly decreased to result in the low pressure injection (LPI) which cooled the heated core. This report summarizes results of the experiment (SB-CL-24) in addition to typical responses of some accident indication systems including the core exit thermocouples (CETs) and the water level meters in the primary system. (author)

Prerouted FPGA Cores for Rapid System Construction in a Dynamic Reconfigurable System

Directory of Open Access Journals (Sweden)

Oliver TimothyF

2007-01-01

Full Text Available A method of constructing prerouted FPGA cores, which lays the foundations for a rapid system construction framework for dynamically reconfigurable computing systems, is presented. Two major challenges are considered: how to manage the wires crossing a core's borders; and how to maintain an acceptable level of flexibility for system construction with only a minimum of overhead. In order to maintain FPGA computing performance, it is crucial to thoroughly analyze the issues at the lowest level of device detail in order to ensure that computing circuit encapsulation is as efficient as possible. We present the first methodology that allows a core to scale its interface bandwidth to the maximum available in a routing channel. Cores can be constructed independently from the rest of the system using a framework that is independent of the method used to place and route primitive components within the core. We use an abstract FPGA model and CAD tools that mirror those used in industry. An academic design flow has been modified to include a wire policy and an interface constraints framework that tightly constrains the use of the wires that cross a core's boundaries. Using this tool set we investigate the effect of prerouting on overall system optimality. Abutting cores are instantly connected by colocation of interface wires. Eliminating run-time routing drastically reduces the time taken to construct a system using a set of cores.

TFTR neutral beam injection system conceptual design

International Nuclear Information System (INIS)

1975-01-01

Three subsystems are described in the following chapters: (1) Neutral Beam Injection Line; (2) Power Supplies; and (3) Controls. Each chapter contains two sections: (1) Functions and Design Requirements; this is a brief listing of the requirements of components of the subsystem. (2) Design Description; this section describes the design and cost estimates. The overall performance requirements of the neutral beam injection system are summarized. (MOW)

Layout considerations for the PSB H- injection system

CERN Document Server

Aiba, M; Carli, C; Chanel, M; Fowler, A; Goddard, B; Weterings, W

2009-01-01

The layout of the PSB H- injection system is described, including the arguments for the geometry and the required equipment performance parameters. The longitudinal positions of the main elements are specified, together with the injected and circulating beam axes. The assumptions used in determining the geometry are listed.

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

Directory of Open Access Journals (Sweden)

A.S. Ekariansyah

2016-08-01

Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

Synthesis of Axial Power Distribution Using 5-Level Ex-core Detector in a Core Protection System

International Nuclear Information System (INIS)

Koo, Bon-Seung; Lee, Chung-Chan; Zee, Sung-Quun

2007-01-01

In ABB-CE digital plants, Core Protection Calculator System (CPCS) is used for a core protection based on several online measured system parameters including 3- level safety grade ex-core detector signals. The CPCS provides four independent channels for the departure from a nucleate boiling ratio (DNBR) and local power density (LPD) trip signals to the reactor protection system. Each channel consists of a core protection calculator (CPC) and a control element assembly calculator (CEAC). The cubic spline synthesis technique has been used in online calculations of the core axial power distributions using 3-level ex-core detector signals in CPC. The pre-determined cubic spline function sets are used depending on the characteristics of the ex-core detector responses. But this method shows large power distribution errors for the extremely skewed axial shapes due to restrictive function sets and an incorrect SAM value. Especially thus situation is worse at a higher burnup. To solve these problems, the cubic spline function sets are improved and it is demonstrated that the axial power shapes can be synthesized more accurately with the new function sets than those of a conventional CPC. In this paper, synthesis of an axial power distribution using a 5-level ex-core detector is described and the axial power distributions are compared between 3-level and 5-level ex-core detector systems

Reflood behavior at low initial clad temperature in Slab Core Test Facility Core-II

International Nuclear Information System (INIS)

Akimoto, Hajime; Sobajima, Makoto; Abe, Yutaka; Iwamura, Takamichi; Ohnuki, Akira; Okubo, Tsutomu; Murao, Yoshio; Okabe, Kazuharu; Adachi, Hiromichi.

1990-07-01

In order to study the reflood behavior with low initial clad temperature, a reflood test was performed using the Slab Core Test Facility (SCTF) with initial clad temperature of 573 K. The test conditions of the test are identical with those of SCTF base case test S2-SH1 (initial clad temperature 1073 K) except the initial clad temperature. Through the comparison of results from these two tests, the following conclusions were obtained. (1) The low initial clad temperature resulted in the low differential pressures through the primary loops due to smaller steam generation in the core. (2) The low initial clad temperature caused the accumulated mass in the core to be increased and the accumulated mass in the downcomer to be decreased in the period of the lower plenum injection with accumulator (before 50s). In the later period of the cold leg injection with LPCI (after 100s), the water accumulation rates in the core and the downcomer were almost the same between both tests. (3) The low initial clad temperature resulted in the increase of the core inlet mass flow rate in the lower plenum injection period. However, the core inlet mass flow rate was almost the same regardless of the initial clad temperature in the later period of the cold leg injection period. (4) The low initial clad temperature resulted in the low turnaround temperature, high temperature rise and fast bottom quench front propagation. (5) In the region apart from the quench front, low initial clad temperature resulted in the lower heat transfer. In the region near the quench front, almost the same heat transfer coefficient was observed between both tests. (6) No flow oscillation with a long period was observed in the SCTF test with low initial clad temperature of 573 K, while it was remarkable in the Cylindrical Core Test Facility (CCTF) test which was performed with the same initial clad temperature. (J.P.N.)

The droplet injection system used in the rod bundle heat transfer facility

International Nuclear Information System (INIS)

Frepoli, C.; Andrew, A.J.; Hochreiter, L.E.; Cheung, F.B.

2001-01-01

The full text follows. The US Nuclear Regulatory Commission (NRC) and the Pennsylvania State University are currently funding a research program entitled ''Rod Bundle Heat Transfer'' (RBHT). The main objective of the program is to investigate heat transfer during the core reflood period of a hypothetical Large Break Loss of Coolant Accident in a typical nuclear power plant. The RBHT test facility consists of a full-length 7 x 7 rod bundle. Information gathered by the RBHT test facility will be used for improvement of the reflood heat transfer models in the NRC's thermal hydraulic codes. In particular the RBHT data will be used to improve the understanding of individual heat transfer effects to the total rod heat transfer such that compensating errors present in current Best Estimate codes can be significantly reduced. The strategy in developing the test matrix is to use a ''building block'' approach in which simpler experiments are performed first to quantify a particular heat transfer mechanism alone and then the additional complications of the full two-phase flow, reflood film boiling behavior of the test facility are added in later experiments. One of these ''simpler'' experiments will be the injection of known size and velocity liquid droplets into the main stream of superheated steam. The droplet injection system consists of small diameter tubes inserted across the bundle at a given elevation. A number of equal size holes are drilled perpendicular to the surface in a triangular pitch. Water is forced into opposite ends of the tube and ejected from the holes. The injection system was tested using a digital imaging system known as VisiSizer. This system is capable of determining the diameter and velocity of small water droplets using a laser-illuminated digital camera system (LIDCS). Imaging software analyzes the digital images in real time to determine the distributions of droplet size and velocity. Pre-test analysis using COBRA-TF have been conducted to

UP-GRADED RHIC INJECTION SYSTEM.

Energy Technology Data Exchange (ETDEWEB)

HAHN,H.FISCHER,W.SEMERTZIDIS,Y.K.WARBURTON,D.S.

2003-05-12

The design of the RHIC injection systems anticipated the possibility of filling and operating the rings with a 120 bunch pattern, corresponding to 110 bunches after allowing for the abort gap. Beam measurements during the 2002 run confirmed the possibility, although at the expense of severe transverse emittance growth and thus not on an operational basis. An improvement program was initiated with the goal of reducing the kicker rise time from 110 to {approx}95 ns and of minimizing pulse timing jitter and drift. The major components of the injection system are 4 kicker magnets and Blmlein pulsers using thyratron switches. The kicker terminating resistor and operating voltage was increased to reduce the rise time. Timing has been stabilized by using commercial trigger units and extremely stable dc supplies for the thyratron reservoir. A fiber optical connection between control room and the thyratron trigger unit has been provided, thereby allowing the operator to adjust timing individually for each kicker unit. The changes were successfully implemented for use in the RHIC operation.

Innovative measurement for injection systems; Innovative Messtechnik fuer Einspritzsysteme

Energy Technology Data Exchange (ETDEWEB)

Janetzky, Bjoern; Majer, Clemens; Doell, Reinahrd [Robert Bosch GmbH, Stuttgart (Germany)

2011-07-01

The treatise introduces a new measuring method to determine the injection quantity and injection rate. In this method the change in pressure within a hydraulic chamber is measured during injection. Also the momentary pressure-dependent speed of sound of the medium inside the chamber is measured. With this information the desired injection quantity can be derived with high precision. A measuring technique (HDA) based on this method is described. Measurement with HDA are performed based on actual requirements from the Diesel-Injection-System, results from these measurements are presented in example. Results are compared with those achieved by using a known measuring technique (EMI). (orig.)

Core cooling systems

International Nuclear Information System (INIS)

Hoeppner, G.

1980-01-01

The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

A benefit assessment of using in-core neutron detector signals in core protection calculator system (CPCS)

International Nuclear Information System (INIS)

Han, S.; Park, S.J.; Seong, P.H.

1997-01-01

A Core Protection Calculator System (CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, in-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piece-wise cubic Spline method is applied; from the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System (COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out. (authors)

A benefit assessment of using in-core neutron detector signals in core protection calculator system(CPCS)

International Nuclear Information System (INIS)

Han, Seung

1996-02-01

A Core Protection Calculator System(CPCS) is a digital computer based safety system generating trip signals based on the calculation of Departure from Nucleate Boiling Ratio(DNBR) and Local Power Density(LPD). Currently, CPCS uses ex-core detector signals for core power calculation and it has some uncertainties. In this study, In-core detector signals which directly measure inside flux of core are applied to CPCS to get more accurate power distribution profile, DNBR and LPD. In order to improve axial power distribution calculation, piecewise cubic spline method is applied: From the 40 nodes of expanded signals, more accurate and detailed core information can be provided. Simulation is carried out to verify its applicability to power distribution calculation. Simulation result shows that the improved method reduces the calculational uncertainties significantly and it allows larger operational margin. It is also expected that no power reduction is required while Core Operating Limit Supervisory System(COLSS) is out-of-service due to reduced uncertainties when the improved method is applied. In this study, a quantitative economic benefit assessment of using in-core neutron detector signals is also carried out

The injection system of the stretcher ring ELSA

International Nuclear Information System (INIS)

Dreist, A.

1989-07-01

For the stretcher ring ELSA in the framwork of this thesis an injection system has been concipated and constructed which should allow all projected operational modes of this stretcher ring, the stretcher, the post-acceleration, and the accumulation mode. The proof could be performed that the realized concept allows all these operational modes. Furthermore it could be shown that the injection shifted from the equilibrium orbit has no disadvantageous effects on a uniform extraction and by this on a high touching ratio. In fact it is even possible to apply the decay of the coherent betatron oscillations around the equilibrium orbit, caused by injection of the incident beam shifted from the equilibrium orbit, to diagnosis purposes: By reproduction of this damping process in a simulation model statements on nonlinearities present in the ring and by this statements on the actual phase-space structure are possible. It has so been shown that the concept presented in this thesis and realized for this thesis represents a suited injection system for the stretcher ring ELSA. (orig.) [de

Proposed high speed pellet injection system 'HIPEL' for Large Helical Device

International Nuclear Information System (INIS)

Sudo, S.; Kanno, M.; Kaneko, H.; Saka, S.; Shirai, T.; Baba, T.

1993-11-01

From the results of the simulation study including pellet ablation and 1-D transport code, it is found that a high speed pellet injector with pellet velocity of more than 3 km/s is necessary for the penetration of the pellet with diameter of 3 mm into the core region under the expected plasma condition of Large Helical Device (LHD) of heliotron/stellarator type with superconducting coils at NIFS in Japan. Therefore, a two stage pellet injector was constructed and tested successfully in order to obtain the pellet velocity range of 3 km/s. Based upon the above results, a high speed flexible multiple-pellet injection system 'HIPEL' for LHD is proposed. HIPEL consists of independent (1) 10 two-stage gun barrels and (2) 10 single-stage gun barrels. It has multi purposes such as refueling and flexible density profile control, diagnostics and the other functions. (author)

An injection system for a linear accelerator

International Nuclear Information System (INIS)

Santos, A.C.R.

1978-03-01

An injection system for the Linear Accelerator is developed using the parameters of machines at the Centro Brasileiro de Pesquisas Fisicas and the Instituto Militar de Engenharia. The proposed system consists basically of a prebuncher and a chopper. The pre-buncher is used to improve the energy resolution and also to increase the accelerator target current. The chopper is used to remove from the beam the electrons that have no possibility of attaining the desired energy and that are usually lost in the walls and the cavity tube, thus producing undesirable background. Theoretical development of the chopper is performed in order to obtain its dimensions for future construction. The complete design the pre-buncher and its feed supply system and the experimental verication of its performance are also presented. It is intended to give the necessary information for the design and construction of the complete injection system proposed. (Author) [pt

A core management system for JRR-3

International Nuclear Information System (INIS)

Soyama, Kazuhiko; Tsuruta, Harumichi; Ichikawa, Hiroki; Nemoto, Hiroyuki.

1991-05-01

Japan Research Reactor No.3 (JRR-3) was upgraded to the thermal output with 20 MW by replacing the core, cooling system and utilization facilities. It is a water moderated and cooled, pool type reactor using 20% enriched U · Alx fuel. A core management system for JRR-3 has been made. This code system can manage of reactivity, power distribution and burn up in consideration of the position of control rod, fuel arrangement and operation pattern. This report is the user's manual of this code system. (author)

Effects of Supercritical Environment on Hydrocarbon-fuel Injection

Institute of Scientific and Technical Information of China (English)

Bongchul Shin; Dohun Kim; Min Son; Jaye Koo

2017-01-01

In this study,the effects of environment conditions on decane were investigated.Decane was injected in subcritical and supercritical ambient conditions.The visualization chamber was pressurized to 1.68 MPa by using nitrogen gas at a temperature of 653 K for subcritical ambient conditions.For supercritical ambient conditions,the visualization chamber was pressurized to 2.52 MPa by using helium at a temperature of 653 K.The decane injection in the pressurized chamber was visualized via a shadowgraph technique and gradient images were obtained by a post processing method.A large variation in density gradient was observed at jet interface in the case of subcritical injection in subcritical ambient conditions.Conversely,for supercritical injection in supercritical ambient conditions,a small density gradient was observed at the jet interface.In a manner similar to that observed in other cases,supercritical injection in subcritical ambient conditions differed from supercritical ambient conditions such as sphere shape liquid.Additionally,there were changes in the interface,and the supercritical injection core width was thicker than that in the subcritical injection.Furthermore,in cases with the same injection conditions,the change in the supercritical ambient normalized core width was smaller than the change in the subcritical ambient normalized core width owing to high specific heat at the supercritical injection and small phase change at the interface.Therefore,the interface was affected by the changing ambient condition.Given that the effect of changing the thermodynamic properties of propellants could be essential for a variable thrust rocket engine,the effects of the ambient conditions were investigated experimentally.

Effects of supercritical environment on hydrocarbon-fuel injection

Science.gov (United States)

Shin, Bongchul; Kim, Dohun; Son, Min; Koo, Jaye

2017-04-01

In this study, the effects of environment conditions on decane were investigated. Decane was injected in subcritical and supercritical ambient conditions. The visualization chamber was pressurized to 1.68 MPa by using nitrogen gas at a temperature of 653 K for subcritical ambient conditions. For supercritical ambient conditions, the visualization chamber was pressurized to 2.52 MPa by using helium at a temperature of 653 K. The decane injection in the pressurized chamber was visualized via a shadowgraph technique and gradient images were obtained by a post processing method. A large variation in density gradient was observed at jet interface in the case of subcritical injection in subcritical ambient conditions. Conversely, for supercritical injection in supercritical ambient conditions, a small density gradient was observed at the jet interface. In a manner similar to that observed in other cases, supercritical injection in subcritical ambient conditions differed from supercritical ambient conditions such as sphere shape liquid. Additionally, there were changes in the interface, and the supercritical injection core width was thicker than that in the subcritical injection. Furthermore, in cases with the same injection conditions, the change in the supercritical ambient normalized core width was smaller than the change in the subcritical ambient normalized core width owing to high specific heat at the supercritical injection and small phase change at the interface. Therefore, the interface was affected by the changing ambient condition. Given that the effect of changing the thermodynamic properties of propellants could be essential for a variable thrust rocket engine, the effects of the ambient conditions were investigated experimentally.

Flow Injection/Sequential Injection Analysis Systems: Potential Use as Tools for Rapid Liver Diseases Biomarker Study

Directory of Open Access Journals (Sweden)

Supaporn Kradtap Hartwell

2012-01-01

Full Text Available Flow injection/sequential injection analysis (FIA/SIA systems are suitable for carrying out automatic wet chemical/biochemical reactions with reduced volume and time consumption. Various parts of the system such as pump, valve, and reactor may be built or adapted from available materials. Therefore the systems can be at lower cost as compared to other instrumentation-based analysis systems. Their applications for determination of biomarkers for liver diseases have been demonstrated in various formats of operation but only a few and limited types of biomarkers have been used as model analytes. This paper summarizes these applications for different types of reactions as a guide for using flow-based systems in more biomarker and/or multibiomarker studies.

Minimally invasive injectable short nanofibers of poly(glycerol sebacate) for cardiac tissue engineering

International Nuclear Information System (INIS)

Ravichandran, Rajeswari; Venugopal, Jayarama Reddy; Sundarrajan, Subramanian; Mukherjee, Shayanti; Sridhar, Radhakrishnan; Ramakrishna, Seeram

2012-01-01

Myocardial tissue lacks the ability to appreciably regenerate itself following myocardial infarction (MI) which ultimately results in heart failure. Current therapies can only retard the progression of disease and hence tissue engineering strategies are required to facilitate the engineering of a suitable biomaterial to repair MI. The aim of this study was to investigate the in vitro properties of an injectable biomaterial for the regeneration of infarcted myocardium. Fabrication of core/shell fibers was by co-axial electrospinning, with poly(glycerol sebacate) (PGS) as core material and poly-l-lactic acid (PLLA) as shell material. The PLLA was removed by treatment of the PGS/PLLA core/shell fibers with DCM:hexane (2:1) to obtain PGS short fibers. These PGS short fibers offer the advantage of providing a minimally invasive injectable technique for the regeneration of infarcted myocardium. The scaffolds were characterized by SEM, FTIR and contact angle and cell–scaffold interactions using cardiomyocytes. The results showed that the cardiac marker proteins actinin, troponin, myosin heavy chain and connexin 43 were expressed more on short PGS fibers compared to PLLA nanofibers. We hypothesized that the injection of cells along with short PGS fibers would increase cell transplant retention and survival within the infarct, compared to the standard cell injection system. (paper)

Superconducting technology for overcurrent limiting in a 25 kA current injection system

Energy Technology Data Exchange (ETDEWEB)

Heydari, Hossein; Faghihi, Faramarz; Sharifi, Reza; Poursoltanmohammadi, Amir Hossein [Center of Excellence for Power System Automation and Operation, Electrical Engineering Department, Iran University of Science and Technology (IUST), Tehran (Iran, Islamic Republic of)], E-mail: heydari@iust.ac.ir, E-mail: faramarz_faghihi@ee.iust.ac.ir, E-mail: reza_sharifi@ee.iust.ac.ir, E-mail: amirhosseinp@ee.iust.ac.ir

2008-09-15

Current injection transformer (CIT) systems are within the major group of the standard type test of high current equipment in the electrical industry, so their performance becomes very important. When designing high current systems, there are many factors to be considered from which their overcurrent protection must be ensured. The output of a CIT is wholly dependent on the impedance of the equipment under test (EUT). Therefore current flow beyond the allowable limit can occur. The present state of the art provides an important guide to developing current limiters not only for the grid application but also in industrial equipment. This paper reports the state of the art in the technology available that could be developed into an application of superconductivity for high current equipment (CIT) protection with no test disruption. This will result in a greater market choice and lower costs for equipment protection solutions, reduced costs and improved system reliability. The paper will also push the state of the art by using two distinctive circuits, closed-core and open-core, for overcurrent protection of a 25 kA CIT system, based on a flux-lock-type superconducting fault current limiter (SFCL) and magnetic properties of high temperature superconducting (HTS) elements. An appropriate location of the HTS element will enhance the rate of limitation with the help of the magnetic field generated by the CIT output busbars. The calculation of the HTS parameters for overcurrent limiting is also performed to suit the required current levels of the CIT.

Superconducting technology for overcurrent limiting in a 25 kA current injection system

Science.gov (United States)

Heydari, Hossein; Faghihi, Faramarz; Sharifi, Reza; Poursoltanmohammadi, Amir Hossein

2008-09-01

Current injection transformer (CIT) systems are within the major group of the standard type test of high current equipment in the electrical industry, so their performance becomes very important. When designing high current systems, there are many factors to be considered from which their overcurrent protection must be ensured. The output of a CIT is wholly dependent on the impedance of the equipment under test (EUT). Therefore current flow beyond the allowable limit can occur. The present state of the art provides an important guide to developing current limiters not only for the grid application but also in industrial equipment. This paper reports the state of the art in the technology available that could be developed into an application of superconductivity for high current equipment (CIT) protection with no test disruption. This will result in a greater market choice and lower costs for equipment protection solutions, reduced costs and improved system reliability. The paper will also push the state of the art by using two distinctive circuits, closed-core and open-core, for overcurrent protection of a 25 kA CIT system, based on a flux-lock-type superconducting fault current limiter (SFCL) and magnetic properties of high temperature superconducting (HTS) elements. An appropriate location of the HTS element will enhance the rate of limitation with the help of the magnetic field generated by the CIT output busbars. The calculation of the HTS parameters for overcurrent limiting is also performed to suit the required current levels of the CIT.

Superconducting technology for overcurrent limiting in a 25 kA current injection system

International Nuclear Information System (INIS)

Heydari, Hossein; Faghihi, Faramarz; Sharifi, Reza; Poursoltanmohammadi, Amir Hossein

2008-01-01

Current injection transformer (CIT) systems are within the major group of the standard type test of high current equipment in the electrical industry, so their performance becomes very important. When designing high current systems, there are many factors to be considered from which their overcurrent protection must be ensured. The output of a CIT is wholly dependent on the impedance of the equipment under test (EUT). Therefore current flow beyond the allowable limit can occur. The present state of the art provides an important guide to developing current limiters not only for the grid application but also in industrial equipment. This paper reports the state of the art in the technology available that could be developed into an application of superconductivity for high current equipment (CIT) protection with no test disruption. This will result in a greater market choice and lower costs for equipment protection solutions, reduced costs and improved system reliability. The paper will also push the state of the art by using two distinctive circuits, closed-core and open-core, for overcurrent protection of a 25 kA CIT system, based on a flux-lock-type superconducting fault current limiter (SFCL) and magnetic properties of high temperature superconducting (HTS) elements. An appropriate location of the HTS element will enhance the rate of limitation with the help of the magnetic field generated by the CIT output busbars. The calculation of the HTS parameters for overcurrent limiting is also performed to suit the required current levels of the CIT

High Power Spark Delivery System Using Hollow Core Kagome Lattice Fibers

Directory of Open Access Journals (Sweden)

Ciprian Dumitrache

2014-08-01

Full Text Available This study examines the use of the recently developed hollow core kagome lattice fibers for delivery of high power laser pulses. Compared to other photonic crystal fibers (PCFs, the hollow core kagome fibers have larger core diameter (~50 µm, which allows for higher energy coupling in the fiber while also maintaining high beam quality at the output (M2 = 1.25. We have conducted a study of the maximum deliverable energy versus laser pulse duration using a Nd:YAG laser at 1064 nm. Pulse energies as high as 30 mJ were transmitted for 30 ns pulse durations. This represents, to our knowledge; the highest laser pulse energy delivered using PCFs. Two fiber damage mechanisms were identified as damage at the fiber input and damage within the bulk of the fiber. Finally, we have demonstrated fiber delivered laser ignition on a single-cylinder gasoline direct injection engine.

An automatic injection system for rapid radiochemistry

International Nuclear Information System (INIS)

Nurmia, M.J.; Kreek, S.A.; Kadkhodayan, B.; Gregorich, K.E.; Lee, D.M.; Hoffman, D.C.

1992-01-01

A description is given of the Automated Injection System (AIS), a pneumatically actuated device for automated collection of nuclear reaction products from a He/KCl gas jet transport system. The AIS is used with the Automated Chemical Chromatographic Element Separation System; together these two devices facilitate completely automated separation procedures with improved speed and reproducibility

Gravity driven emergency core cooling experiments with the PACTEL facility

International Nuclear Information System (INIS)

Munther, R.; Kalli, H.; Kouhia, J.

1996-01-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs

Gravity driven emergency core cooling experiments with the PACTEL facility

Energy Technology Data Exchange (ETDEWEB)

Munther, R; Kalli, H [University of Technology, Lappeenranta (Finland); Kouhia, J [Technical Research Centre of Finland, Lappeenranta (Finland)

1996-12-01

PACTEL (Parallel Channel Test Loop) is an experimental out-of-pile facility designed to simulated the major components and system behaviour of a commercial Pressurized Water Reactor (PWR) during different postulated LOCAs and transients. The reference reactor to the PACTEL facility is Loviisa type WWER-440. The recently made modifications enable experiments to be conducted also on the passive core cooling. In these experiments the passive core cooling system consisted of one core makeup tank (CMT) and pressure balancing lines from the pressurizer and from a cold leg connected to the top of the CMT in order to maintain the tank in pressure equilibrium with the primary system during ECC injection. The line from the pressurizer to the core makeup tank was normally open. The ECC flow was provided from the CMT located at a higher elevation than the main part of the primary system. A total number of nine experiments have been performed by now. 4 refs, 7 figs, 3 tabs.

Armor systems including coated core materials

Science.gov (United States)

Chu, Henry S [Idaho Falls, ID; Lillo, Thomas M [Idaho Falls, ID; McHugh, Kevin M [Idaho Falls, ID

2012-07-31

An armor system and method involves providing a core material and a stream of atomized coating material that comprises a liquid fraction and a solid fraction. An initial layer is deposited on the core material by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is less than the liquid fraction of the stream of atomized coating material on a weight basis. An outer layer is then deposited on the initial layer by positioning the core material in the stream of atomized coating material wherein the solid fraction of the stream of atomized coating material is greater than the liquid fraction of the stream of atomized coating material on a weight basis.

First order study for an iron core OH system for TNS

International Nuclear Information System (INIS)

Ballou, J.K.; Schultz, J.

1977-01-01

A simple comparison has been made between an air core and an iron core ohmic heating system for a particular device, and it was shown that the peak power requirements can be substantially reduced by the use of an iron core to power levels handled by industry today. It was also shown that for an ohmic heating system initiated plasma that the cost of the iron core ohmic heating power system (iron core, dual rectifier, and DC switch) is less than the cost for a subset of the power system for an air core system (dual rectifier and DC switch). There is considerable work being done on other methods of initiating the plasma none of which seem to be incompatible with the use of an iron core system

Uncovering the information core in recommender systems

Science.gov (United States)

Zeng, Wei; Zeng, An; Liu, Hao; Shang, Ming-Sheng; Zhou, Tao

2014-08-01

With the rapid growth of the Internet and overwhelming amount of information that people are confronted with, recommender systems have been developed to effectively support users' decision-making process in online systems. So far, much attention has been paid to designing new recommendation algorithms and improving existent ones. However, few works considered the different contributions from different users to the performance of a recommender system. Such studies can help us improve the recommendation efficiency by excluding irrelevant users. In this paper, we argue that in each online system there exists a group of core users who carry most of the information for recommendation. With them, the recommender systems can already generate satisfactory recommendation. Our core user extraction method enables the recommender systems to achieve 90% of the accuracy of the top-L recommendation by taking only 20% of the users into account. A detailed investigation reveals that these core users are not necessarily the large-degree users. Moreover, they tend to select high quality objects and their selections are well diversified.

Thermal protection system for the concrete core support floor at Fort St. Vrain

International Nuclear Information System (INIS)

Jones, H.; Hedgecock, P.D.

1976-01-01

A unique feature of the Fort St. Vrain HTGR is its steel jacketed concrete core support floor. The construction of this floor generally resembles that of the prestressed concrete reactor vessel, but its location immediately below the core hot gas outlets generates some particularly severe thermal protection requirements. A thermal barrier is used over the entire outer surface of the floor and in the twelve hot gas ducts which convey the primary coolant through the floor to the steam generators. A cooling water system of square tubes welded to the inside of the steel jacket is used to remove that heat which does pass through the thermal barrier and to maintain the concrete at acceptable temperatures. The design approach to the floor itself and to the thermal barriers and cooling system will be described, but the main emphasis of the paper will be on the total experience gained during construction and pre-operational testing. A particular problem experienced during construction was leakage from some cooling tubes, after their embedment in concrete. The solution to that problem was to develop a method for injecting catalyzed epoxy into the leaking tube. This method, which has general usefulness for in-service repairs, will be described. (author)

Core supervision methods and future improvements of the core master/presto system at KKB

International Nuclear Information System (INIS)

Lundberg, S.; Wenisch, J.; Teeffelen, W.V.

2000-01-01

Kernkraftwerk Brunsbuettel (KKB) is a KWU 806 MW e BWR located at the lower river Elbe, in Germany. The reactor has been in operation since 1976 and is now operating in its 14. cycle. The core supervision at KKB is performed with the ABB CORE MASTER system. This system mainly contains the 3-D simulator PRESTO supplied by Studsvik Scandpower A/S. The core supervision is performed by periodic PRESTO 3-D evaluations of the reactor operation state. The power distribution calculated by PRESTO is adapted with the ABB UPDAT program using the on-line LPRM readings. The thermal margins are based on this adapted power distribution. Related to core supervision, the function of the PRESTO/UPDAT codes is presented. The UPDAT method is working well and is capable of reproducing the true core power distribution. The quality of the 3-D calculation is, however, an important ingredient of the quality of the adapted power distribution. The adaptation method as such is also important for this quality. The data quality of this system during steady state and off-rate states (reactor manoeuvres) are discussed by presenting comparisons between PRESTO and UPDAT thermal margin utilisation from Cycle 13. Recently analysed asymmetries in the UPDAT evaluated MCPR values are also presented and discussed. Improvements in the core supervision such as the introduction of advanced modern nodal methods (PRESTO-2) are presented and an alternative core supervision philosophy is discussed. An ongoing project with the goal to update the data and result presentation interface (GUI) is also presented. (authors)

Improvement of JRR-4 core management code system

International Nuclear Information System (INIS)

Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N.

2000-01-01

In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

A direct plasma injection system into an RFQ for clean and safe ion implantation

International Nuclear Information System (INIS)

Takeuchi, T.; Katayama, T.; Okamura, M.; Yano, K.; Sakumi, A.; Hattori, T.; Hayashizaki, N.; Jameson, R.A.

2002-01-01

A new injection system, direct plasma injection system, was tested and its principle was proved successfully. We found that one of advantages of this injection system was efficient consumption of source materials. Large portions of induced ions can be injected into a first stage accelerator. This feature is quite useful for ion implantation applications, because toxic exhaust gas can be eliminated. In order to utilize this system for industrial application, the feasibility of a boron injection scheme using a Nd:YAG laser system was investigated

Design and development of a direct injection system for cryogenic engines

Science.gov (United States)

Mutumba, Angela; Cheeseman, Kevin; Clarke, Henry; Wen, Dongsheng

2018-04-01

The cryogenic engine has received increasing attention due to its promising potential as a zero-emission engine. In this study, a new robust liquid nitrogen injection system was commissioned and set up to perform high-pressure injections into an open vessel. The system is used for quasi-steady flow tests used for the characterisation of the direct injection process for cryogenic engines. An electro-hydraulic valve actuator provides intricate control of the valve lift, with a minimum cycle time of 3 ms and a frequency of up to 20 Hz. With additional sub-cooling, liquid phase injections from 14 to 94 bar were achieved. Results showed an increase in the injected mass with the increase in pressure, and decrease in temperature. The injected mass was also observed to increases linearly with the valve lift. Better control of the injection process, minimises the number of variables, providing more comparable and repeatable sets of data. Implications of the results on the engine performance were also discussed.

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

International Nuclear Information System (INIS)

Scheuerer, Martina; Weis, Johannes

2012-01-01

Highlights: ► Pressurized thermal shocks are important phenomena for plant life extension and aging. ► The thermal-hydraulics of PTS have been studied experimentally and numerically. ► In the Large Scale Test Facility a loss of coolant accident was investigated. ► CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

The PEP II injection kicker system

International Nuclear Information System (INIS)

Pappas, G.C.; Donaldson, A.R.; Williams, D.

1997-07-01

PEP II or the B Factory consists of two asymmetric storage rings. The injection energy for electrons is 9 GeV, while that for positrons is 3.1 GeV. The bend angle into the high energy ring (HER) is 0.35 m-rad, and the angle into the low energy ring (LER) is 0.575 m-rad. The magnetic length for the HER kicker is 0.85 m, and 0.55 m for the LER kicker. The field produced by the magnet is therefore 123.5 G for the HER, and 132 G for the LER. Each ring has a kicker magnet upstream of the injection line which is used to distort the orbit of the stored beam. An identical magnet downstream of the injection line is used to restore the orbit of the stored beam and inject the incoming beam. The two magnets are driven in parallel by the modulator. The apeture of the magnets is 3.86x3.46 cm (HxV). Therefore the current required to drive the HER is 863 A, while for the LER it is 756 A. The inductance of the magnet is approximately 1.4 uH/m. The current pulse is a critically damped sinusoid with a rise time of less than 300 ns. A kicker system has been designed which can be used for injection of both beams by varying the charge of voltage. The modulator uses a conjugate circuit to match the impedance of the magnet, and coupling to the beam chamber

Cooling water injection system

International Nuclear Information System (INIS)

Inai, Nobuhiko.

1989-01-01

In a BWR type reactor, ECCS system is constituted as a so-called stand-by system which is not used during usual operation and there is a significant discontinuity in relation with the usual system. It is extremely important that ECCS operates upon occurrence of accidents just as specified. In view of the above in the present invention, the stand-by system is disposed along the same line with the usual system. That is, a driving water supply pump for supplying driving water to a jet pump is driven by a driving mechanism. The driving mechanism drives continuously the driving water supply pump in a case if an expected accident such as loss of the function of the water supply pump, as well as during normal operation. That is, all of the water supply pump, jet pump, driving water supply pump and driving mechanism therefor are caused to operate also during normal operation. The operation of them are not initiated upon accident. Thus, the cooling water injection system can perform at high reliability to remarkably improve the plant safety. (K.M.)

Availability analysis of the AP600 passive core cooling system

Energy Technology Data Exchange (ETDEWEB)

Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

1996-12-01

The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, Jozef; Vocka, Radim

2010-01-01

The SCORPIO (SCORPIO-VVER) core monitoring system, its basic features and history of implementation at Czech NPPs are described. The most important improvements in the area of neutron physics, core thermal analysis and operation support are as follows: Moving to the 42 axial nodes across the whole system (2004); Implementation of new cross section library to support mixed reactor core with differences in axial geometry of used fuel types and enhancement of Core Simulator boundary conditions model, to properly address the 'wild' geometry in axial direction; Adjusting the thermohydraulic and neutron-physical models regarding to the Gd2 fuel needs; Support up to 5 types of FAs and 2 types of SPND (Posit, IST); Extension of form functions for pin-wise reconstruction to improve pin-power prediction in control rod coupler region; System adaptation to the new upgraded digital I and C unit system; Integration of the SCORPIO-VVER system and its workstation into the plant redundant in-core system; Implementation of new On-Line form function generation to module RECON; New design of the Strategy Generator with advanced predictions; Adaptation of the system to support the new up-rated reactor thermal power; Adding new online SDM calculation function into to system; Implementation of the new 3D power reconstruction with SPND interpretation; Extending the limit checking to the 'full core' checking. The control of margins to the technical specification: Extended to full core - all FA is controlled individually in core; The limits are definable up to 59 FA (1/6 symmetry); 4 limited parameters are controlled - Kr, qlin, Tout-fa, dTfa; 2 additional parameters are monitored - dTsat, DNBR; New MMI are developed to present the limited and controlled parameters in core. Upgrade 3 is planned for the Slovak Bohunice NPP in 2011-2012. (P.A.)

Development of In-Core Protection System

International Nuclear Information System (INIS)

Cho, J. H; Kim, C. H.; Kim, J. H.; Jeong, S. H.; Sohn, S. D.; BaeK, S. M.; YOON, J. H.

2016-01-01

In-core Protection System (ICOPS) is an on-line digital computer system which continuously calculates Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) based on plant parameters to make trip decisions based on the computations. The function of the system is the same as that of Core Protection Calculator System (CPCS) and Reactor Core Protection System (RCOPS) which are applied to Optimized Power Reactor 1000 (OPR1000) and Advanced Power Reactor 1400 (APR1400). The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. To achieve this goal, several algorithms were newly developed and hardware and software design was updated. The functional design requirements document was developed by KEPCO-NF and the component design was conducted by Doosan. System design and software implementation were performed by KEPCO-E and C, and software Verification and Validation (V and V) was performed by KEPCO-E and C and Sure Softtech. The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. The function of I/O simulator was improved even though the hardware platform is the same as that of RCOPS for Shin-Hanul 1 and 2. SCADE was applied to the implementation of ICOPS software, and the V and V system for ICOPS which satisfies international standards was developed. Although several further detailed design works remain, the function of ICOPS has been confirmed. The ICOPS will be applied to APR+ project, and the further works will be performed in following project

Development of In-Core Protection System

Energy Technology Data Exchange (ETDEWEB)

Cho, J. H; Kim, C. H.; Kim, J. H.; Jeong, S. H.; Sohn, S. D.; BaeK, S. M.; YOON, J. H. [KEPCO Engineering and Construction Co., Deajeon (Korea, Republic of)

2016-10-15

In-core Protection System (ICOPS) is an on-line digital computer system which continuously calculates Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD) based on plant parameters to make trip decisions based on the computations. The function of the system is the same as that of Core Protection Calculator System (CPCS) and Reactor Core Protection System (RCOPS) which are applied to Optimized Power Reactor 1000 (OPR1000) and Advanced Power Reactor 1400 (APR1400). The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. To achieve this goal, several algorithms were newly developed and hardware and software design was updated. The functional design requirements document was developed by KEPCO-NF and the component design was conducted by Doosan. System design and software implementation were performed by KEPCO-E and C, and software Verification and Validation (V and V) was performed by KEPCO-E and C and Sure Softtech. The ICOPS has been developed to overcome the algorithm related obstacles in overseas project. The function of I/O simulator was improved even though the hardware platform is the same as that of RCOPS for Shin-Hanul 1 and 2. SCADE was applied to the implementation of ICOPS software, and the V and V system for ICOPS which satisfies international standards was developed. Although several further detailed design works remain, the function of ICOPS has been confirmed. The ICOPS will be applied to APR+ project, and the further works will be performed in following project.

The Influence of runner system on production of injection molds

Directory of Open Access Journals (Sweden)

Janostik Vaclav

2016-01-01

Full Text Available This experimental study describes the influence of runner system on rheological properties during the injection molding process. Economic effects on the amount of production are discussed as well. Autodesk Moldflow Synergy 2016 (Moldflow was used for the study of the injection process. Three suggestions of the runner system, cold runner system, hot runner system and the combination of cold–hot runner system have been promoted. These three variants underwent the rheological and economic analysis. As a result, recommendations for the application of the runner system for the required amount of production have been suggested

Experimental Study and Mathematical Modeling of Asphaltene Deposition Mechanism in Core Samples

Directory of Open Access Journals (Sweden)

Jafari Behbahani T.

2015-11-01

Full Text Available In this work, experimental studies were conducted to determine the effect of asphaltene deposition on the permeability reduction and porosity reduction of carbonate, sandstone and dolomite rock samples using an Iranian bottom hole live oil sample which is close to reservoir conditions, whereas in the majority of previous work, a mixture of recombined oil (a mixture of dead oil and associated gas was injected into a core sample which is far from reservoir conditions. The effect of the oil injection rate on asphaltene deposition and permeability reduction was studied. The experimental results showed that an increase in the oil injection flow rate can result in an increase in asphaltene deposition and permeability reduction. Also, it can be observed that at lower injection flow rates, a monotonic decrease in permeability of the rock samples can be attained upon increasing the injection flow rate, while at higher injection rates, after a decrease in rock permeability, an increasing trend is observed before a steady-state condition can be reached. The experimental results also showed that the rock type can affect the amount of asphaltene deposition, and the asphaltene deposition has different mechanisms in sandstone and carbonate core samples. It can be seen that the adsorption and plugging mechanisms have a more important role in asphaltene deposition in carbonate core samples than sandstone core samples. From the results, it can be observed that the pore volumes of the injected crude oil are higher for sandstone cores compared with the carbonate cores. Also, it can be inferred that three depositional types may take place during the crude oil injection, i.e., continuous deposition for low-permeability cores, slow, steady plugging for high-permeability cores and steady deposition for medium-permeability cores. It can be seen from the experimental results that damage to the core samples was found to increase when the production pressures were

Development of lab scale fast gas injection system for SST-1 Tokamak

International Nuclear Information System (INIS)

Pathan, F.S.; Banaudha, Moni; Khristi, Yohan; Khan, M.S.; Khan, Ziauddin; Raval, D.C.; Khirwadkar, Samir

2017-01-01

The plasma density control plays an important role in Tokamak operation. The factors that influence plasma density in a Tokamak device are working gas injection, pumping, ionization rate and the recycle coefficient representing the wall conditions. Among these factors, gas injection is relatively convenient to be controlled. Hence, the most frequently adopted method to control the plasma density is to control the fast gas injection. This paper describes the design and experimental work carried out towards the development of Fast Gas Injection System for SST-1 Tokamak. Laboratory based test setup was successfully established for Fast Gas Injection System that can feed predefined quantity of gas in a controlled manner into vacuum chamber. Further, this FGIS system will be implemented in SST-1 Tokamak environment with online density feedback signal

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

Development of integrated control system for smart factory in the injection molding process

Science.gov (United States)

Chung, M. J.; Kim, C. Y.

2018-03-01

In this study, we proposed integrated control system for automation of injection molding process required for construction of smart factory. The injection molding process consists of heating, tool close, injection, cooling, tool open, and take-out. Take-out robot controller, image processing module, and process data acquisition interface module are developed and assembled to integrated control system. By adoption of integrated control system, the injection molding process can be simplified and the cost for construction of smart factory can be inexpensive.

Development of self-forming doxorubicin-loaded polymeric depots as an injectable drug delivery system for liver cancer chemotherapy.

Science.gov (United States)

Nittayacharn, Pinunta; Nasongkla, Norased

2017-07-01

The objective of this work was to develop self-forming doxorubicin-loaded polymeric depots as an injectable drug delivery system for liver cancer chemotherapy and studied the release profiles of doxorubicin (Dox) from different depot formulations. Tri-block copolymers of poly(ε-caprolactone), poly(D,L-lactide) and poly(ethylene glycol) named PLECs were successfully used as a biodegradable material to encapsulate Dox as the injectable local drug delivery system. Depot formation and encapsulation efficiency of these depots were evaluated. Results show that depots could be formed and encapsulate Dox with high drug loading content. For the release study, drug loading content (10, 15 and 20% w/w) and polymer concentration (25, 30, and 35% w/v) were varied. It could be observed that the burst release occurred within 1-2 days and this burst release could be reduced by physical mixing of hydroxypropyl-beta-cyclodextrin (HP-β-CD) into the depot system. The degradation at the surface and cross-section of the depots were examined by Scanning Electron Microscope (SEM). In addition, cytotoxicity of Dox-loaded depots and blank depots were tested against human liver cancer cell lines (HepG2). Dox released from depots significantly exhibited potent cytotoxic effect against HepG2 cell line compared to that of blank depots. Results from this study reveals an important insight in the development of injectable drug delivery system for liver cancer chemotherapy. Schematic diagram of self-forming doxorubicin-loaded polymeric depots as an injectable drug delivery system and in vitro characterizations. (a) Dox-loaded PLEC depots could be formed with more than 90% of sustained-release Dox at 25% polymer concentration and 20% Dox-loading content. The burst release occurred within 1-2 days and could be reduced by physical mixing of hydroxypropyl-beta-cyclodextrin (HP-β-CD) into the depot system. (b) Dox released from depots significantly exhibited potent cytotoxic effect against human

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Braun, H. E.; Dollard, W. J.; Tower, S. N.

1980-01-01

A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

Numerical modeling of injection, stress and permeability enhancement during shear stimulation at the Desert Peak Enhanced Geothermal System

Science.gov (United States)

Dempsey, David; Kelkar, Sharad; Davatzes, Nick; Hickman, Stephen H.; Moos, Daniel

2015-01-01

Creation of an Enhanced Geothermal System relies on stimulation of fracture permeability through self-propping shear failure that creates a complex fracture network with high surface area for efficient heat transfer. In 2010, shear stimulation was carried out in well 27-15 at Desert Peak geothermal field, Nevada, by injecting cold water at pressure less than the minimum principal stress. An order-of-magnitude improvement in well injectivity was recorded. Here, we describe a numerical model that accounts for injection-induced stress changes and permeability enhancement during this stimulation. In a two-part study, we use the coupled thermo-hydrological-mechanical simulator FEHM to: (i) construct a wellbore model for non-steady bottom-hole temperature and pressure conditions during the injection, and (ii) apply these pressures and temperatures as a source term in a numerical model of the stimulation. In this model, a Mohr-Coulomb failure criterion and empirical fracture permeability is developed to describe permeability evolution of the fractured rock. The numerical model is calibrated using laboratory measurements of material properties on representative core samples and wellhead records of injection pressure and mass flow during the shear stimulation. The model captures both the absence of stimulation at low wellhead pressure (WHP ≤1.7 and ≤2.4 MPa) as well as the timing and magnitude of injectivity rise at medium WHP (3.1 MPa). Results indicate that thermoelastic effects near the wellbore and the associated non-local stresses further from the well combine to propagate a failure front away from the injection well. Elevated WHP promotes failure, increases the injection rate, and cools the wellbore; however, as the overpressure drops off with distance, thermal and non-local stresses play an ongoing role in promoting shear failure at increasing distance from the well.

[Systemic safety following intravitreal injections of anti-VEGF].

Science.gov (United States)

Baillif, S; Levy, B; Girmens, J-F; Dumas, S; Tadayoni, R

2018-03-01

The goal of this manuscript is to assess data suggesting that intravitreal injection of anti-vascular endothelial growth factors (anti-VEGFs) could result in systemic adverse events (AEs). The class-specific systemic AEs should be similar to those encountered in cancer trials. The most frequent AE observed in oncology, hypertension and proteinuria, should thus be the most common expected in ophthalmology, but their severity should be lower because of the much lower doses of anti-VEGFs administered intravitreally. Such AEs have not been frequently reported in ophthalmology trials. In addition, pharmacokinetic and pharmacodynamic data describing systemic diffusion of anti-VEGFs should be interpreted with caution because of significant inconsistencies reported. Thus, safety data reported in ophthalmology trials and pharmacokinetic/pharmacodynamic data provide robust evidence that systemic events after intravitreal injection are very unlikely. Additional studies are needed to explore this issue further, as much remains to be understood about local and systemic side effects of anti-VEGFs. Copyright © 2018 Elsevier Masson SAS. All rights reserved.

Study on core make-up water experiment of AC600 make-up water tank

International Nuclear Information System (INIS)

Ji Fuyun; Li Changlin; Zheng Hua; Liu Shaohua; Xu Xiaolan

1999-01-01

The core makeup tank (CMT) is a principal component of the passive high pressure safety injection systems for AC600 and has a function to inject cold borated water into reactor vessel during abnormal events. The purpose of this experiment is to verify the gravity drain behavior of the CMT and to provide experimental data to verify the computer codes used in the safety analyses. Five experiments with simulative small and medium break conditions are conducted at AC600 core makeup tank performance test facility of Nuclear Power Institute of China (NPIC). The author provides the results of one test. The simulated accident is a small break loss-of-coolant accident

Low-Power Embedded DSP Core for Communication Systems

Science.gov (United States)

Tsao, Ya-Lan; Chen, Wei-Hao; Tan, Ming Hsuan; Lin, Maw-Ching; Jou, Shyh-Jye

2003-12-01

This paper proposes a parameterized digital signal processor (DSP) core for an embedded digital signal processing system designed to achieve demodulation/synchronization with better performance and flexibility. The features of this DSP core include parameterized data path, dual MAC unit, subword MAC, and optional function-specific blocks for accelerating communication system modulation operations. This DSP core also has a low-power structure, which includes the gray-code addressing mode, pipeline sharing, and advanced hardware looping. Users can select the parameters and special functional blocks based on the character of their applications and then generating a DSP core. The DSP core has been implemented via a cell-based design method using a synthesizable Verilog code with TSMC 0.35[InlineEquation not available: see fulltext.]m SPQM and 0.25[InlineEquation not available: see fulltext.]m 1P5M library. The equivalent gate count of the core area without memory is approximately 50 k. Moreover, the maximum operating frequency of a[InlineEquation not available: see fulltext.] version is 100 MHz (0.35[InlineEquation not available: see fulltext.]m) and 140 MHz (0.25[InlineEquation not available: see fulltext.]m).

Transient computational fluid dynamics analysis of emergency core cooling injection at natural circulation conditions

Energy Technology Data Exchange (ETDEWEB)

Scheuerer, Martina, E-mail: Martina.Scheuerer@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany); Weis, Johannes, E-mail: Johannes.Weis@grs.de [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Forschungsinstitute, 85748 Garching (Germany)

2012-12-15

Highlights: Black-Right-Pointing-Pointer Pressurized thermal shocks are important phenomena for plant life extension and aging. Black-Right-Pointing-Pointer The thermal-hydraulics of PTS have been studied experimentally and numerically. Black-Right-Pointing-Pointer In the Large Scale Test Facility a loss of coolant accident was investigated. Black-Right-Pointing-Pointer CFD software is validated to simulate the buoyancy driven flow after ECC injection. - Abstract: Within the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), computational fluid dynamics (CFD) software is validated for the simulation of the thermo-hydraulics of pressurized thermal shocks. A proposed validation experiment is the test series performed within the OECD ROSA V project in the Large Scale Test Facility (LSTF). The LSTF is a 1:48 volume-scaled model of a four-loop Westinghouse pressurized water reactor (PWR). ROSA V Test 1-1 investigates temperature stratification under natural circulation conditions. This paper describes calculations which were performed with the ANSYS CFD software for emergency core cooling injection into one loop at single-phase flow conditions. Following the OECD/NEA CFD Best Practice Guidelines (Mahaffy, 2007) the influence of grid resolution, discretisation schemes, and turbulence models (shear stress transport and Reynolds stress model) on the mixing in the cold leg were investigated. A half-model was used for these simulations. The transient calculations were started from a steady-state solution at natural circulation conditions. The final calculations were obtained in a complete model of the downcomer. The results are in good agreement with data.

Injections of the selective adenosine A2A antagonist MSX-3 into the nucleus accumbens core attenuate the locomotor suppression induced by haloperidol in rats.

Science.gov (United States)

Ishiwari, Keita; Madson, Lisa J; Farrar, Andrew M; Mingote, Susana M; Valenta, John P; DiGianvittorio, Michael D; Frank, Lauren E; Correa, Merce; Hockemeyer, Jörg; Müller, Christa; Salamone, John D

2007-03-28

There is considerable evidence of interactions between adenosine A2A receptors and dopamine D2 receptors in striatal areas, and antagonists of the A2A receptor have been shown to reverse the motor effects of DA antagonists in animal models. The D2 antagonist haloperidol produces parkinsonism in humans, and also induces motor effects in rats, such as suppression of locomotion. The present experiments were conducted to study the ability of the adenosine A2A antagonist MSX-3 to reverse the locomotor effects of acute or subchronic administration of haloperidol in rats. Systemic (i.p.) injections of MSX-3 (2.5-10.0 mg/kg) were capable of attenuating the suppression of locomotion induced by either acute or repeated (i.e., 14 day) administration of 0.5 mg/kg haloperidol. Bilateral infusions of MSX-3 directly into the nucleus accumbens core (2.5 microg or 5.0 microg in 0.5 microl per side) produced a dose-related increase in locomotor activity in rats treated with 0.5 mg/kg haloperidol either acutely or repeatedly. There were no overall significant effects of MSX-3 infused directly into the dorsomedial nucleus accumbens shell or the ventrolateral neostriatum. These results indicate that antagonism of adenosine A2A receptors can attenuate the locomotor suppression produced by DA antagonism, and that this effect may be at least partially mediated by A2A receptors in the nucleus accumbens core. These studies suggest that adenosine and dopamine systems interact to modulate the locomotor and behavioral activation functions of nucleus accumbens core.

Railgun pellet injection system for fusion experimental devices

International Nuclear Information System (INIS)

Onozuka, M.; Hasegawa, K.

1995-01-01

A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s -1 using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s -1 using a 3 m long railgun. (orig.)

Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

International Nuclear Information System (INIS)

Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

1977-02-01

Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

Depth protection system

International Nuclear Information System (INIS)

Arita, Setsuo; Izumi, Shigeru; Suzuki, Satoru; Noguchi, Atomi.

1988-01-01

Purpose: To previously set a nuclear reactor toward safety side by the reactor scram if an emergency core cooling system is failed to operate. Constitution If abnormality occurs in an emergency core cooling system or an aqueous boric acid injection system, a reactor protection system is operated and, if the reactor protection system shows an abnormal state, a control rod withdrawal inhibition system is operated as a fundamental way. For instance, when the driving power source voltage for the emergency core cooling system is detected and, if it is lower than a predetermined value, the reactor protection system is operated. Alternatively, if the voltage goes lower than the predetermined value, the control rod withdrawal is inhibited. In addition, stopping for the feedwater system is inhibited. Further, integrity of the driving means for the emergency core cooling system is positively checked and the protection function is operated depending on the result of check. Since the nuclear reactor can be set toward the safety side even if the voltage for the driving power source of the aqueous boric acid injection system is lower than a predetermined value, the reactor safety can further be improved. (Horiuchi, T.)

Fast Flux Test Facility core restraint system performance

International Nuclear Information System (INIS)

Hecht, S.L.; Trenchard, R.G.

1990-02-01

Characterizing Fast Flux Test Facility (FFTF) core restraint system performance has been ongoing since the first operating cycle. Characterization consists of prerun analysis for each core load, in-reactor and postirradiation measurements of subassembly withdrawal loads and deformations, and using measurement data to fine tune predictive models. Monitoring FFTF operations and performing trend analysis has made it possible to gain insight into core restraint system performance and head off refueling difficulties while maximizing component lifetimes. Additionally, valuable information for improved designs and operating methods has been obtained. Focus is on past operating experience, emphasizing performance improvements and avoidance of potential problems. 4 refs., 12 figs., 2 tabs

Systemic barriers accessing HIV treatment among people who inject drugs in Russia: a qualitative study.

Science.gov (United States)

Sarang, Anya; Rhodes, Tim; Sheon, Nicolas

2013-10-01

Achieving 'universal access' to antiretroviral HIV treatment (ART) in lower income and transitional settings is a global target. Yet, access to ART is shaped by local social condition and is by no means universal. Qualitative studies are ideally suited to describing how access to ART is socially situated. We explored systemic barriers to accessing ART among people who inject drugs (PWID) in a Russian city (Ekaterinburg) with a large burden of HIV treatment demand. We undertook 42 in-depth qualitative interviews with people living with HIV with current or recent experience of injecting drug use. Accounts were analysed thematically, and supplemented here with an illustrative case study. Three core themes were identified: 'labyrinthine bureaucracy' governing access to ART; a 'system Catch 22' created by an expectation that access to ART was conditional upon treated drug use in a setting of limited drug treatment opportunity; and 'system verticalization', where a lack of integration across HIV, tuberculosis (TB) and drug treatment compromised access to ART. Taken together, we find that systemic factors play a key role in shaping access to ART with the potential adverse effects of reproducing treatment initiation delay and disengagement from treatment. We argue that meso-level systemic factors affecting access to ART for PWID interact with wider macro-level structural forces, including those related to drug treatment policy and the social marginalization of PWID. We note the urgent need for systemic and structural changes to improve access to ART for PWID in this setting, including to simplify bureaucratic procedures, foster integrated HIV, TB and drug treatment services, and advocate for drug treatment policy reform.

Injection system for the PEP II asymmetric B Factory at SLAC

International Nuclear Information System (INIS)

Fieguth, T.; Bloom, E.; Bulos, F.; Davies-White, W.; Donald, M.; Fairfield, K.; Godfrey, G.; Holtzapple, R.; Ronan, M.; Zisman, M.

1992-01-01

The asymmetric energy B Factory proposed as an upgrade of PEP at Stanford Linear Accelerator Center will require a highly reliable and efficient injection system. The conceptual design has shown the feasibility of extracting 9 GeV electrons and 3.1 GeV positrons from the existing linac and injecting equal charges into 1658 buckets in each of the two rings of the collider. An injection study group has continued the development and study of this proposal and has generated workable designs for many related systems and subsystems. (author) 3 refs.; 4 figs.; 2 tabs

Resolution of issues related to alternative RCS injection in the absence of containment sump recirculation

International Nuclear Information System (INIS)

Charles L Kling; Stephen S Barshay; Mathew C Jacob; Michael J Friedman

2005-01-01

Full text of publication follows: On June 9, 2003 the US NRC issued Bulletin No. 2003-01 that deals with the potential impact of debris blockage on containment sump recirculation at PWRs during a Loss-of-Coolant Accident (LOCA). In response to the bulletin, the Omaha Public Power District (OPPD) is in the process of developing procedural and operational strategies for their Fort Calhoun Station (FCS) to address the issues raised. Westinghouse provided engineering support to OPPD in identifying and resolving issues related to alternative means of supplying safety injection water to the reactor coolant system (RCS) in the absence of containment sump recirculation. Nuclear power plants are designed to protect the core following a LOCA by providing a continuous supply of cooling water to the core. In the long term, the Refueling Water Storage Tank (RWST) inventory will be depleted and core heat removal accomplished via recirculation of water previously injected into the Reactor Coolant System (RCS) and collected in the containment sump. Debris generated within the containment as a result of the impingement of fluid jets in the Zone of Influence (ZOI) of the RCS break and containment wash down may find its way into the containment sump. As the safety injection pumps take suction from the sump, in the recirculation mode of operation, the debris suspended in the sump water could begin to accumulate in the sump screen that is located in the recirculation path. Should sufficient debris accumulate on the sump screen, a flow blockage could potentially develop. This would result in insufficient safety injection pump NPSH, thereby impairing the recirculation mode of injection into RCS. Potential debris blockage and prevention of sump recirculation may be addressed by refilling the RWST with water and injecting this water directly into the core. This paper identifies and attempts to resolve several issues related to this alternative mode of RCS injection. In particular, the

Hydrogen ion species analysis and related neutral beam injection power assessment in the Heliotron E neutral beam injection system

International Nuclear Information System (INIS)

Sano, Fumimichi; Obiki, Tokuhiro; Sasaki, Akihiko; Iiyoshi, Atsuo; Uo, Koji

1982-01-01

The hydrogen ion species in a Heliotron E neutral beam injection system of maximum electric power 6.3 MW were analyzed in order to assess the neutral beam power injected into the torus. The masimum p roton ratio of the cylindrical bucket type ion source used was observed to be more than 90 percent assuming that the angular divergences for the respective species in the beam are the same. The experimental data are compared with calculations using a particle balance model. The analysis indicates that the net injection power reaches nearly 2.7 MW at the optimal conditions of the system considering the geometrical limitation of the neutral beam path. (author)

Data archiving system implementation in ITER's CODAC Core System

International Nuclear Information System (INIS)

Castro, R.; Abadie, L.; Makushok, Y.; Ruiz, M.; Sanz, D.; Vega, J.; Faig, J.; Román-Pérez, G.; Simrock, S.; Makijarvi, P.

2015-01-01

Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

An experimental study of injection and spray characteristics of diesel and gasoline blends on a common rail injection system

International Nuclear Information System (INIS)

Han, Dong; Wang, Chunhai; Duan, Yaozong; Tian, Zhisong; Huang, Zhen

2014-01-01

The injection and spray characteristics of diesel and gasoline blends are investigated on a common rail injection system. The injection rate, fuel spray evolution process (tip penetration distance, spray cone angle, projected spray area and relative brightness intensity contour) and microscopic droplet features are analyzed. The results show that diesel and gasoline blends have higher volumetric injection rates, earlier starts of injection and shorter injection delays, but little variances are observed in the mass injection rates for different test fuels. Increased gasoline proportion in the test blends causes slightly decreased spray tip penetration distance but increased spray cone angle. Also, more smaller-size droplets are observed in the fuel jet of the diesel and gasoline blends, indicating that the spray breakup and atomization processes are promoted. - Highlights: • Injection rate and spray characteristics of diesel and gasoline blends are studied. • Diesel and gasoline blends have higher volumetric injection rates. • Earlier starts of injection are found when using diesel and gasoline blends. • Diesel and gasoline blends produce shorter spray penetration but higher cone angle. • The number of small droplets increases in the spray of diesel and gasoline blends

Immunological Risk of Injectable Drug Delivery Systems

NARCIS (Netherlands)

Jiskoot, W.; van Schie, R.M.F.; Carstens, M.G.; Schellekens, H.

2009-01-01

Injectable drug delivery systems (DDS) such as particulate carriers and water-soluble polymers are being used and developed for a wide variety of therapeutic applications. However, a number of immunological risks with serious clinical implications are associated with administration of DDS. These

WGA-Alexa transsynaptic labeling in the phrenic motor system of adult rats: Intrapleural injection versus intradiaphragmatic injection.

Science.gov (United States)

Buttry, Janelle L; Goshgarian, Harry G

2015-02-15

Intrapleural injection of CTB-Alexa 488, a retrograde tracer, provides an alternative labeling technique to the surgically invasive laparotomy required for intradiaphragmatic injection. However, CTB-Alexa 488 is incapable of crossing synapses restricting the tracer to the phrenic nuclei and the intercostal motor nuclei in the spinal cord. Intrapleural injection of WGA-Alexa 488, a transsynaptic tracer, provides a method to label the respiratory motor pathway in both the spinal cord and medulla. Intradiaphragmatic injection of WGA-Alexa 594 and vagal nerve injections of True blue were used to confirm the phrenic nuclei and to differentiate between the rVRG and the NA in the medulla. Following intrapleural injection, WGA-Alexa 488 was retrogradely transported to the phrenic nuclei and to the intercostal motor nuclei. Subsequently WGA-Alexa 488 was transsynaptically transported from the phrenic motoneurons to the pre-motor neurons in the rVRG that provide the descending drive to the phrenic neurons during inspiration. In addition WGA-Alexa 488 was identified in select cells of the NA confirmed by a dual label of both WGA-Alexa 488 and True blue. WGA-Alexa 488 demonstrates retrograde transsynaptic labeling following intrapleural injection whereas the previous method of injecting CTB-Alexa 488 only demonstrates retrograde labeling. Intrapleural injection of WGA-Alexa fluor conjugates is an effective method to transsynaptically label the phrenic motor system providing an alternative for the invasive laparotomy required for intradiaphragmatic injections. Furthermore, the study provides the first anatomical evidence of a direct synaptic relationship between rVRG and select NA cells. Copyright © 2014 Elsevier B.V. All rights reserved.

Railgun pellet injection system for fusion experimental devices

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Oda, Y. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Azuma, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Satake, K. [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Adv. Tech. Dev. Dept.; Kasai, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan); Hasegawa, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun 319-11 (Japan)

1995-11-01

A railgun pellet injection system has been developed for fusion experimental devices. Using a low electric energy railgun system, hydrogen pellet acceleration tests have been conducted to investigate the application of the electromagnetic railgun system for high speed pellet injection into fusion plasmas. In the system, the pellet is pre-accelerated before railgun acceleration. A laser beam is used to induce plasma armature. The ignited plasma armature is accelerated by an electromagnetic force that accelerates the pellet. Under the same operational conditions, the energy conversion coefficient for the dummy pellets was around 0.4%, while that for the hydrogen pellets was around 0.12%. The highest hydrogen pellet velocity was 1.4 km s{sup -1} using a 1 m long railgun. Based on the findings, it is estimated that the hydrogen pellet has the potential to be accelerated to 5 km s{sup -1} using a 3 m long railgun. (orig.).

Polymeric microchip for the simultaneous determination of anions and cations by hydrodynamic injection using a dual-channel sequential injection microchip electrophoresis system.

Science.gov (United States)

Gaudry, Adam J; Nai, Yi Heng; Guijt, Rosanne M; Breadmore, Michael C

2014-04-01

A dual-channel sequential injection microchip capillary electrophoresis system with pressure-driven injection is demonstrated for simultaneous separations of anions and cations from a single sample. The poly(methyl methacrylate) (PMMA) microchips feature integral in-plane contactless conductivity detection electrodes. A novel, hydrodynamic "split-injection" method utilizes background electrolyte (BGE) sheathing to gate the sample flows, while control over the injection volume is achieved by balancing hydrodynamic resistances using external hydrodynamic resistors. Injection is realized by a unique flow-through interface, allowing for automated, continuous sampling for sequential injection analysis by microchip electrophoresis. The developed system was very robust, with individual microchips used for up to 2000 analyses with lifetimes limited by irreversible blockages of the microchannels. The unique dual-channel geometry was demonstrated by the simultaneous separation of three cations and three anions in individual microchannels in under 40 s with limits of detection (LODs) ranging from 1.5 to 24 μM. From a series of 100 sequential injections the %RSDs were determined for every fifth run, resulting in %RSDs for migration times that ranged from 0.3 to 0.7 (n = 20) and 2.3 to 4.5 for peak area (n = 20). This system offers low LODs and a high degree of reproducibility and robustness while the hydrodynamic injection eliminates electrokinetic bias during injection, making it attractive for a wide range of rapid, sensitive, and quantitative online analytical applications.

Restraint system for core elements of a reactor core

International Nuclear Information System (INIS)

Class, G.

1975-01-01

In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

Initial operation and performance of the PDX neutral-beam injection system

International Nuclear Information System (INIS)

Kugel, H.W.; Eubank, H.P.; Kozub, T.A.; Rossmassler, J.E.; Schilling, G.; van Halle, A.; Williams, M.D.

1982-01-01

In 1981, the joint ORNL/PPPL PDX neutral beam heating project succeeded in reliably injecting 7.2 MW of D 0 into the PDX plasma, at nearly perpendicular angles, and achieved ion temperatures up to 6.5 keV. The expeditious achievement of this result was due to the thorough conditioning and qualification of the PDX neutral beam ion sources at ORNL prior to delivery coupled with several field design changes and improvements in the injection system made at PPPL as a result of neutral beam operating experience with the PLT tokamak. It has been found that the operation of high power neutral beam injection systems in a tokamak-neutral beam environment requires procedures and performance different from those required for development operation on test stands. In this paper, we review the installatin of the PDX neutral beam injection system, and its operation and performance during the initial high power plasma heating experiments with the PDX tokamak

LMFBR core flowering response to an impulse load

International Nuclear Information System (INIS)

Brochard, D.; Petret, J.C.; Queval, J.C.; Gibert, R.J.

1993-01-01

Some incidental situations like MFCI (Meeting Fuel Coolant Incident) may induce a core flowering and lead to consider impulse loans applied to LMFBR core. These highly dynamic loads are very different considering their spatial repartition and their frequency content from the seismic loads which have been deeply studied. Recently, tests have been performed on the LMFBR core mock-up RAPSODIE in order to validate the calculation methods for centered impulse load. These tests consist in injecting water quickly in the mock-up through a specific device replacing the core central assembly. The influence of the injection pressure and the influence of the injection axial position have been investigate. During the tests, the top displacements of some assemblies have been measured. The aim of this paper is first to present the experimental device and the test results. Then a non linear numerical model is described; this model includes the impact between subassemblies and is based on an homogenization method allowing to take into account with accuracy the fluid structure interaction.The comparisons between calculation results an test results will finally be presented

Voltage controlled nano-injection system for single-cell surgery

Science.gov (United States)

Seger, R. Adam; Actis, Paolo; Penfold, Catherine; Maalouf, Michelle; Vilozny, Boaz; Pourmand, Nader

2015-01-01

Manipulation and analysis of single cells is the next frontier in understanding processes that control the function and fate of cells. Herein we describe a single-cell injection platform based on nanopipettes. The system uses scanning microscopy techniques to detect cell surfaces, and voltage pulses to deliver molecules into individual cells. As a proof of concept, we injected adherent mammalian cells with fluorescent dyes. PMID:22899383

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

International Nuclear Information System (INIS)

Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed

Preliminary Uncertainty Analysis for SMART Digital Core Protection and Monitoring System

International Nuclear Information System (INIS)

Koo, Bon Seung; In, Wang Kee; Hwang, Dae Hyun

2012-01-01

The Korea Atomic Energy Research Institute (KAERI) developed on-line digital core protection and monitoring systems, called SCOPS and SCOMS as a part of SMART plant protection and monitoring system. SCOPS simplified the protection system by directly connecting the four RSPT signals to each core protection channel and eliminated the control element assembly calculator (CEAC) hardware. SCOMS adopted DPCM3D method in synthesizing core power distribution instead of Fourier expansion method being used in conventional PWRs. The DPCM3D method produces a synthetic 3-D power distribution by coupling a neutronics code and measured in-core detector signals. The overall uncertainty analysis methodology which is used statistically combining uncertainty components of SMART core protection and monitoring system was developed. In this paper, preliminary overall uncertainty factors for SCOPS/SCOMS of SMART initial core were evaluated by applying newly developed uncertainty analysis method

Emulation study on system characteristic of high pressure common-rail fuel injection system for marine medium-speed diesel engine

Science.gov (United States)

Wang, Qinpeng; Yang, Jianguo; Xin, Dong; He, Yuhai; Yu, Yonghua

2018-05-01

In this paper, based on the characteristic analyzing of the mechanical fuel injection system for the marine medium-speed diesel engine, a sectional high-pressure common rail fuel injection system is designed, rated condition rail pressure of which is 160MPa. The system simulation model is built and the performance of the high pressure common rail fuel injection system is analyzed, research results provide the technical foundation for the system engineering development.

Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

International Nuclear Information System (INIS)

Nawathe, Shirish; Umashankari, P.; Balakrishnan, Kamala; Mahajan, S.C.; Kakodkar, A.

1991-01-01

A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

Energy Technology Data Exchange (ETDEWEB)

Tuunanen, J. [VTT Energy, Espoo (Finland). Nuclear Energy; Vihavainen, J. [Lappeenranta Univ. of Technology (Finland); D' Auria, F. [Univ. of Pisa (Italy); Kimber, G. [AEA Technology (United Kingdom)

1999-07-01

The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

International Nuclear Information System (INIS)

Tuunanen, J.; D'Auria, F.; Kimber, G.

1999-01-01

The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

An on-line adaptive core monitoring system

International Nuclear Information System (INIS)

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

1997-01-01

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

Experimentation of a fixed in-core-based system for core limiting conditions of operation (LCO) monitoring

International Nuclear Information System (INIS)

Piguet, F.; Carrasco, M.; Mourlevat, J.L.; Rio, G.; Verneret, C.

2006-01-01

In order to comply with the needs of Utilities for improvements in the economic competitiveness of nuclear energy, one of the solutions proposed is to reduce the cost of the fuel cycle. To this aim, increasing the lifetime of cycles by introducing so-called 'low leakage' fuel loading patterns to the reactor is a rather promising solution. However, these loading patterns lead to an increase in the core hotspot factors and therefore to a reduction in the operating margins with respect to the core operating limits also called 'Limiting Conditions of Operations (LCO)'. For many years FRAMATOME-ANP has developed and proposed solutions aiming at increasing and therefore restoring these margins, namely: the improvement in design methods based on three-dimensional modelling of the core, on kinetic representation of transients and on neutron-thermohydraulic coupling or the improvement in the fuel with the introduction of intermediate grids. A complementary approach is to improve the core instrumentation associated with the system for monitoring the core operating margins to the LCO thresholds. The core operating limits monitoring function calls on real-time knowledge of the current power distribution in the core. If we take the French 1300 MWe units as an example, this knowledge is based on the measurement of the mean axial power distribution made by six sections neutron detectors, located outside the pressure vessel and equipped with a fast neutron filtering device. The results of this measurement are combined with pre-tabulated radial hotspot factors (Fxy), in order to calculate the total hotspot factor (FQ) of the core, the minimum Departure from Nucleate Boiling Ratio (DNBR) and, consequently, the margins with respect to the core operating limits. The limitations of a measurement made outside the vessel, and those of the 1D/2D modelling adopted, mean that these margins calculations have a high potential for improving the level of their accuracy. This is the reason why

Novel Double-Needle System That Can Prevent Intravascular Injection of Any Filler

Directory of Open Access Journals (Sweden)

Hsiang Huang, MD

2017-09-01

Full Text Available Summary:. A new type of needle system combines 2 parts, an inner needle and an outer needle. The inner needle is used for filler injection and the outer needle acts as a guiding needle that can observe blood reflow when inserting into the vessel lumen during injection process. This new needle system can be used for all kinds of filler, providing real time monitoring for physician and preventing intravascular injection of any filler.

Core Flight System Satellite Starter Kit

Data.gov (United States)

National Aeronautics and Space Administration — The Core Flight System Satellite Starter Kit (cFS Kit) will allow a small satellite or CubeSat developer to rapidly develop, deploy, test, and operate flight...

Injection characteristics of dimethyl ether

Energy Technology Data Exchange (ETDEWEB)

Glensvig, M.

1996-09-01

Dimethyl ether (DME) has proved to be a new ultra-clean alternative fuel for diesel engines. Engine tests have shown considerably lower NO{sub x} emissions, no particle emissions and lower noise compared to that obtained from normal diesel engine operation. DME also has demonstrated favorable response to Exhaust Gas Recirculation (EGR). The purpose of this investigation was to achieve a better understanding of the fundamental spray behavior of DME. Fundamental spray behaviour was characterized by fuel spray penetration and angle, atomization and droplet size and evaporation. The influence of fuel characteristics, nozzle geometry and ambient pressure on the DME and diesel spray behavior was investigated. Fuel was injected into an unheated injection chamber with a ambient pressure of 15 bar and 25 bar, respectively, giving a simplified simulation of the environment in an operating engine. Two nozzles were studied: a single hole nozzle and a pintle nozzle. A conventional fuel injection system was used for both nozzles. Injection parameters of RPM, throttle position, fuel line length and chamber environment were held constant for both nozzles. The sprays were visualized using schlieren and high speed photography. Results show that the general appearance of the DME spray is similar to that of diesel spray. The core of the DME spray seems less dense and the spray tip less sharp compared to diesel spray, indicating smaller droplets with a lower momentum in the core of the DME spray. Schlieren film shows that with both DME and diesel fuel, the spray tip only consists of liquid and that evaporation occurs after a brief time interval. Penetration of DME is about one third that of diesel using the pintle nozzle. Also, the spray angle is considerably larger for the DME spray compared to the diesel spray. A comparatively smaller difference in penetration is observed using the hole nozzle. Differences in penetration for the hole nozzle are within the limit of the penetration

Effects of upper plenum injection on thermo-hydrodynamic behavior under refill and reflood phases

International Nuclear Information System (INIS)

Iwamura, Takamichi; Sobajima, Makoto; Abe, Yutaka; Adachi, Hiromichi; Ohnuki, Akira; Osakabe, Masahiro

1984-12-01

In order to investigate the thermo-hydrodynamic behavior in core under simultaneous ECC water injection into the upper plenum and the intact cold leg during the refill and reflood phases of a PWR-LOCA, Tests S1-SH3 and S1-SH4 were performed by using Slab Core Test Facility (SCTF) with the injection of saturated and 67K subcooled water into the upper plenum, respectively, under the same cold leg injection condition. The following major findings were obtained by examining these test results. (1) Although the core was cooled by the fall back water from the upper plenum into the core during the period of high injection rate into the upper plenum, the core was cooled mainly by the bottom flooding after the BOCREC (Bottom of core recovery). (2) The possible fall back flow rate estimated with a CCFL correlation rapidly decreased after the BOCREC because of the increase of steam generation rate in core. (3) Continuous fall back of subcooled water was not observed even under the condition with large upper plenum injection rate of subcooled water and with steam outflow through the lower plenum into the downcomer. The fall back was intermittently limited by the rapid increase of upward steam flow which was generated in the core due to the evaporation of the fall back water. (4) The rising of liquid level in the lower plenum was suppressed by the pressurization in core due to the evaporation of fall back water before the BOCREC and therefore the beginning of bottom reflood was delayed. Some selected data from Tests S1-SH3 and S1-SH4 are also included in this report. (author)

SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

International Nuclear Information System (INIS)

Molnar, J.; Vocka, R.

2010-01-01

In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

Conceptual design of a neutral-beam injection system for the TFTR

International Nuclear Information System (INIS)

Ehlers, K.W.; Berkner, K.H.; Cooper, W.S.; Hooper, E.B.; Pyle, R.V.; Stearns, J.W.

1975-11-01

The neutral-beam injection requirements for heating and fueling the next generation of fusion reactor experiments far exceed those of present devices; the neutral-beam systems needed to meet these requirements will be large and complex. A conceptual design of a TFTR tokamak injection system to produce 120 keV deuterium-ion beams with a total power of about 80 MW is given

Overview of the design of core restraint systems

International Nuclear Information System (INIS)

Heinecke, J.

1984-01-01

The optimization of the core restraint system is an important condition for the safe and reliable operation of a fast breeder reactor. For KNK II which is under successful operation and SNR 300 all requirements for safety and operation have been met with help of a ring type system. For SNR 2 the decision between the ring type system and the free standing core has to be done in the near future. Within these considerations the advantages of a ring type restraint system of limiting deflections during operation and limiting of possible movements under seismic conditions have to be balanced against the somewhat more complicated structure of the ring type restraint system

Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

1995-09-01

Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

An expert system for PWR core operation management

Energy Technology Data Exchange (ETDEWEB)

Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa

1988-01-01

Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers.

An expert system for PWR core operation management

International Nuclear Information System (INIS)

Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa.

1988-01-01

Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers

Emergency core cooling device

International Nuclear Information System (INIS)

Suzaki, Kiyoshi; Inoue, Akihiro.

1979-01-01

Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

Plasma core reactor applications

International Nuclear Information System (INIS)

Latham, T.S.; Rodgers, R.J.

1976-01-01

Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

Control of Surge in Centrifugal Compressor by Using a Nozzle Injection System: Universality in Optimal Position of Injection Nozzle

Directory of Open Access Journals (Sweden)

Toshiyuki Hirano

2012-01-01

Full Text Available The passive control method for surge and rotating stall in centrifugal compressors by using a nozzle injection system was proposed to extend the stable operating range to the low flow rate. A part of the flow at the scroll outlet of a compressor was recirculated to an injection nozzle installed on the inner wall of the suction pipe of the compressor through the bypass pipe and injected to the impeller inlet. Two types of compressors were tested at the rotational speeds of 50,000 rpm and 60,000 rpm with the parameter of the circumferential position of the injection nozzle. The present experimental results revealed that the optimum circumferential position, which most effectively reduced the flow rate for the surge inception, existed at the opposite side of the tongue of the scroll against the rotational axis and did not depend on the compressor system and the rotational speeds.

Advanced gadolinia core and Toshiba advanced reactor management system

International Nuclear Information System (INIS)

Miyamoto, Toshiki; Yoshioka, Ritsuo; Ebisuya, Mitsuo

1988-01-01

At the Hamaoka Nuclear Power Station, Unit No. 3, advanced core design and core management technology have been adopted, significantly improving plant availability, operability and reliability. The outstanding technologies are the advanced gadolinia core (AGC) which utilizes gadolinium for the axial power distribution control, and Toshiba advanced reactor management system (TARMS) which uses a three-dimensional core physics simulator to calculate the power distribution. Presented here are the effects of these advanced technologies as observed during field testing. (author)

Relativistic electron beam source with an air-core step-up transformer

International Nuclear Information System (INIS)

Mohri, Akihiro; Ikuta, Kazunari; Masuzaki, Masaru; Tsuzuki, Tetsuya; Fujiwaka, Setsuya.

1975-04-01

An air-core step-up transformer with a high coupling factor has been developed to generate a high voltage pulse for charging the pulse forming line of a relativistic electron beam source. A beam source using the transformer was constructed and well operated for the beam injection into a toroidal system. (auth.)

Linear stability analysis of the gas injection augmented natural circulation of STAR-LM

International Nuclear Information System (INIS)

Yeon-Jong Yoo; Qiao Wu; James J Sienicki

2005-01-01

Full text of publication follows: A linear stability analysis has been performed for the gas injection augmented natural circulation of the Secure Transportable Autonomous Reactor - Liquid Metal (STAR-LM). Natural circulation is of great interest for the development of Generation-IV nuclear energy systems due to its vital role in the area of passive safety and reliability. One of such systems is STAR-LM under development by Argonne National Laboratory. STAR-LM is a 400 MWt class modular, proliferation-resistant, and passively safe liquid metal-cooled fast reactor system that uses inert lead (Pb) coolant and the advanced power conversion system that consists of a gas turbine Brayton cycle utilizing supercritical carbon dioxide (CO 2 ) to obtain higher plant efficiency. The primary loop of STAR-LM relies only on the natural circulation to eliminate the use of circulation pumps for passive safety consideration. To enhance the natural circulation of the primary coolant, STAR-LM optionally incorporates the additional driving force provided by the injection of noncondensable gas into the primary coolant above the reactor core, which is effective in removing heat from the core and transferring it to the secondary working fluid without the attainment of excessive coolant temperature at nominal operating power. Therefore, it naturally raises the concern about the natural circulation instability due to the relatively high temperature change in the core and the two-phase flow condition in the hot leg above the core. For the ease of analysis, the flow path of the loop was partitioned into five thermal-hydraulically distinct sections, i.e., heated core, unheated core, hot leg, heat exchanger, and cold leg. The one-dimensional single-phase flow field equations governing the natural circulation, i.e., continuity, momentum, and energy equations, were used for each section except the hot leg. For the hot leg, the one-dimensional homogeneous equilibrium two-phase flow field

Computer based core monitoring system for an operating CANDU reactor

International Nuclear Information System (INIS)

Yoon, Moon Young; Kwon, O Hwan; Kim, Kyung Hwa; Yeom, Choong Sub

2004-01-01

The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

Environmental response nanosilica for reducing the pressure of water injection in ultra-low permeability reservoirs

Science.gov (United States)

Liu, Peisong; Niu, Liyong; Li, Xiaohong; Zhang, Zhijun

2017-12-01

The super-hydrophobic silica nanoparticles are applied to alter the wettability of rock surface from water-wet to oil-wet. The aim of this is to reduce injection pressure so as to enhance water injection efficiency in low permeability reservoirs. Therefore, a new type of environmentally responsive nanosilica (denote as ERS) is modified with organic compound containing hydrophobic groups and "pinning" groups by covalent bond and then covered with a layer of hydrophilic organic compound by chemical adsorption to achieve excellent water dispersibility. Resultant ERS is homogeneously dispersed in water with a size of about 4-8 nm like a micro-emulsion system and can be easily injected into the macro or nano channels of ultra-low permeability reservoirs. The hydrophobic nanosilica core can be released from the aqueous delivery system owing to its strong dependence on the environmental variation from normal condition to injection wells (such as pH and salinity). Then the exposed silica nanoparticles form a thin layer on the surface of narrow pore throat, leading to the wettability from water-wet to oil-wet. More importantly, the two rock cores with different permeability were surface treated with ERS dispersion with a concentration of 2 g/L, exhibit great reduce of water injection pressure by 57.4 and 39.6%, respectively, which shows great potential for exploitation of crude oil from ultra-low permeability reservoirs during water flooding. [Figure not available: see fulltext.

Study of RF system of Hefei storage ring under injection

International Nuclear Information System (INIS)

Xu Hongliang; Wang Lin; Li Yongjun; Huang Guirong; Zhang Pengfei; Li Weimin; Liu Zuping; He Duohui

2004-01-01

In this paper, the beam loading effect of RF system and the conditions of Robinson instability are analyzed in detail. By the study of the injection beam intensity limit dependent on detune angle and visible detune angle, it is found that the storage ring can be injected to more than 300 mA current intensity to attain the design target of phase II project in the lower energy injection situation of Hefei Storage Ring if a certain power is feed in the RF cavity and a certain tuning angle of the RF cavity is set

Applications of plasma core reactors to terrestrial energy systems

International Nuclear Information System (INIS)

Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

AA injection kicker in its tank

CERN Multimedia

CERN PhotoLab

1980-01-01

For single-turn injection of the antiprotons, a septum at the end of the injection line made the beam parallel to the injection orbit, and a quarter of a betatron-wavelength downstream a fast kicker corrected the angle. Kicker type: lumped delay line. PFN voltage 56 kV. Bending angle 7.5 mrad; kick-strength 0.9 Tm; fall-time 95%-5% in 150 ns. The injection orbit is to the left, the stack orbit to the far right. A fast shutter near the central orbit had to be closed before the kicker fired, so as to protect the stack core from being shaken by the kicker's fringe field. The shutter is shown in closed position.

Safety evaluation report on Westinghouse Electric Company ECCS evaluation model for plants equipped with upper head injection

International Nuclear Information System (INIS)

Lauben, G.N.; Wagner, N.H.; Israel, S.L.; McPherson, G.D.; Hodges, M.W.

1978-04-01

For plants which include an ice condenser containment concept, Westinghouse has planned an additional safety system known as the upper head injection (UHI) system to augment the emergency core cooling system. This system is comprised of additional accumulator tanks and piping arranged to supply cooling water to the top of the core during the blowdown period following a postulated large-break loss-of-coolant accident (LOCA). The objective of UHI is to add to the core cooling provided by the conventional emergency core cooling system (ECCS) and so permit operation at linear heat rates comparable to those permitted in plants utilizing the dry containment concept. In this way, plants which include the UHI system would have greater operating flexibility while still meeting the acceptance criteria as defined in paragraph 50.46 of 10 CFR Part 50. This review is concerned with those changes to the Westinghouse ECCS evaluation model that have been proposed for the UHI-LOCA model

Simulation of the aspersion system of the core at high pressure (HPCS) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Vargas O, D.; Chavez M, C.

2012-10-01

A high-priority topic for the nuclear industry is the safety, consequently a nuclear power plant should have the emergency systems of cooling of the core (ECCS), designed exclusively to enter in operation in the event of an accident with coolant loss, including the design base accident. The objective of the aspersion system of the core at high pressure (HPCS) is to provide in an autonomous way the cooling to the core maintaining for if same the coolant inventory even when a small break is presented that does not allow the depressurization of the reactor and also avoiding excessive temperatures that affect the shielding of the fuel. The present work describes the development of the model and the simulation of the HPCS using the RELAP/SCDAP code. During the process simulation, for the setting in march of the system HPCS in an accident with coolant loss is necessary to implement the main components of the system taking into account what unites them, the main pump, the filled pump, the suction and injection valves, pipes and its water sources that can be condensed storage tanks and the suppression pool. The simulation of this system will complement the model with which counts the Analysis Laboratory in Nuclear Reactors Engineering of the UNAM regarding to the nuclear power plant of Laguna Verde which does not have a detailed simulation of the emergency cooling systems. (Author)

A Common Definition of the System Operators' Core Activities

International Nuclear Information System (INIS)

2006-02-01

In this report a common definition of the system operator's core activities in the Nordic countries is identified and also a list of non-core activities is introduced. As a starting point the common tasks for system responsibility as identified by Nordel has been used for the work. The term TSO (Transmission System Operator) is employed as a common denominator in the report. It is found out that the TSOs carry out common core activities in the roles as a transmission operator, a system operator and a balance settlement responsible. The core activities for the TSO as a transmission network operator are: Maintain the adequate transmission system in the long run and network development plan on the national as well as on the Nordic level using sophisticated analysis and planning methods and tools. Plan the transmission network on the national as well as on the Nordic level utilising new investments, renewal and maintenance of existing network components so that the network is secure to operate and adequate transmission capacity is guaranteed. Aim at timely network expansions using enhanced information exchange between the Nordic TSOs, and on the national level between the TSO and distribution and regional network operators, large consumers and large producers. Secure the technical compatibility with networks across the border and within a country by establishing connection requirements on the national level and ensuring that the national requirements are compatible across the Nordic power system. The core activities for the TSO as a system operator are: Define common technical requirements for the secure system operation using common planning, operation, connection and data exchange procedures. Secure the system operation with the operational planning for the following year by using information exchange between TSOs enabling the TSOs to make the best possible forecast of the global grid situation in order to assess the flows in their network and the available

Core design characteristics of the hyper system

International Nuclear Information System (INIS)

Yonghee, Kim; Won-Seok, Park; Hill, R.N.

2003-01-01

In Korea, an accelerator-driven system (ADS) called HYPER (Hybrid Power Extraction Reactor) is being studied for the transmutation of the radioactive wastes. HYPER is a 1000 MWth lead-bismuth eutectic (LBE)-cooled ADS. In this paper, the neutronic design characteristics of HYPER are described and its transmutation performances are assessed for an equilibrium cycle. The core is loaded with a ductless fuel assembly containing transuranics (TRU) dispersion fuel pins. In HYPER, a relatively high core height, 160 cm, is adopted to maximize the multiplication efficiency of the external source. In the ductless fuel assembly, 13 non-fuel rods are used as tie rods to maintain the mechanical integrity of assembly. As the reflector material, pure lead is used to improve the neutron economy and to minimise the generation of radioactive materials. In HYPER, to minimise the burn-up reactivity swing, a B 4 C burnable absorber is employed. For efficient depletion of the B-10 absorber, the burnable absorber is loaded only in the axially-central part (92 cm long) of the 13 tie rods of each assembly. In the current design, the amount of the B 4 C absorber was determined such that the burn-up reactivity swing is about 3.0% Δk. The long-lived fission products (LLFPs) 99 Tc and 129 I are also transmuted in the HYPER core such that their supporting ratios are equal to that of the TRUs. A heterogeneous LLFP transmutation in the reflector zone has been analysed in this work. A unique feature of the HYPER system is that it has an auxiliary core shutdown system, independent of the accelerator shutdown system. It has been shown that a cylindrical B 4 C absorber between the target and fuel blanket can drastically reduce the fission power even without shutting off the accelerator power. (author)

Development of whole core thermal-hydraulic analysis program ACT. 3. Coupling core module with primary heat transport system module

International Nuclear Information System (INIS)

Ohtaka, Masahiko; Ohshima, Hiroyuki

1998-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including inter-wrapper flow under various reactor operation conditions. In this work, the core module as a main part of the ACT developed last year, which simulates thermal-hydraulics in the subassemblies and the inter-subassembly gaps, was coupled with an one dimensional plant system thermal-hydraulic analysis code LEDHER to simulate transients in the primary heat transport system and to give appropriate boundary conditions to the core model. The effective algorithm to couple these two calculation modules was developed, which required minimum modification of them. In order to couple these two calculation modules on the computing system, parallel computing technique using PVM (Parallel Virtual Machine) programming environment was applied. The code system was applied to analyze an out-of-pile sodium experiment simulating core with 7 subassemblies under transient condition for code verification. It was confirmed that the analytical results show a similar tendency of experimental results. (author)

Real-time simulation of ex-core nuclear instrumentation system

International Nuclear Information System (INIS)

Zhao Qiang; Zhang Zhijian; Cao Xinrong

2005-01-01

Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

International Nuclear Information System (INIS)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

2017-01-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Fault tree analysis on BWR core spray system

International Nuclear Information System (INIS)

Watanabe, Norio

1982-06-01

Fault Trees which describe the failure modes for the Core Spray System function in the Browns Ferry Nuclear Plant (BWR 1065MWe) were developed qualitatively and quantitatively. The unavailability for the Core Spray System was estimated to be 1.2 x 10 - 3 /demand. It was found that the miscalibration of four reactor pressure sensors or the failure to open of the two inboard valves (FCV 75-25 and 75-53) could reduce system reliability significantly. It was recommended that the pressure sensors would be calibrated independently. The introduction of the redundant inboard valves could improve the system reliability. Thus this analysis method was verified useful for system analysis. The detailed test and maintenance manual and the informations on the control logic circuits of each active component are necessary for further analysis. (author)

Injection control development of the JT-60U electron cyclotron heating system

Energy Technology Data Exchange (ETDEWEB)

Hiranai, Shinichi; Shinozaki, Shin-ichi; Yokokura, Kenji; Moriyama, Shinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Fumiaki [Nippon Advanced Technology Co., Ltd., Tokai, Ibaraki (Japan); Suzuki, Yasuo [Atomic Energy General Service Co., Ltd., Tokai, Ibaraki (Japan); Ikeda, Yoshitaka [Japan Atomic Energy Research Inst., Kashiwa, Chiba (Japan)

2003-03-01

The JT-60U electron cyclotron heating (ECH) System injects a millimeteric wave at 110 GHz into the JT-60 Plasma, and heats the plasma or drives a current locally to enhance the confinement performance of the JT-60 plasma. The system consists of four sets of high power gyrotrons, high voltage power supplies and transmission lines, and two antennas that launch electron cyclotron (EC) beams toward the plasma. The key features of the injection control system are streering of the direction of the EC beam by driving the movable mirror in the antenna, and capability to set any combination of polarization angle and ellipticity by rotating the two grooved mirrors in the polarizers. This report represents the design, fabrication and improvements of the injection control system. (author)

Development of hybrid core calculation system using 2-D full-core heterogeneous transport calculation and 3-D advanced nodal calculation

International Nuclear Information System (INIS)

Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi

2004-01-01

This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)

Power injected in dissipative systems and the fluctuation theorem

Science.gov (United States)

Aumaître, S.; Fauve, S.; McNamara, S.; Poggi, P.

We consider three examples of dissipative dynamical systems involving many degrees of freedom, driven far from equilibrium by a constant or time dependent forcing. We study the statistical properties of the injected and dissipated power as well as the fluctuations of the total energy of these systems. The three systems under consideration are: a shell model of turbulence, a gas of hard spheres colliding inelastically and excited by a vibrating piston, and a Burridge-Knopoff spring-block model. Although they involve different types of forcing and dissipation, we show that the statistics of the injected power obey the ``fluctuation theorem" demonstrated in the case of time reversible dissipative systems maintained at constant total energy, or in the case of some stochastic processes. Although this may be only a consequence of the theory of large deviations, this allows a possible definition of ``temperature" for a dissipative system out of equilibrium. We consider how this ``temperature" scales with the energy and the number of degrees of freedom in the different systems under consideration.

Application of startup/core management code system to YGN 3 startup testing

International Nuclear Information System (INIS)

Chi, Sung Goo; Hah, Yung Joon; Doo, Jin Yong; Kim, Dae Kyum

1995-01-01

YGN 3 is the first nuclear power plant in Korea to use the fixed incore detector system for startup testing and core management. The startup/core management code system was developed from existing ABB-C-E codes and applied for YGN 3 startup testing, especially for physics and CPC(Core Protection Calculator)/COLSS (Core Operating Limit Supervisory System) related testing. The startup/core management code system consists of startup codes which include the CEBASE, CECOR, CEFAST and CEDOPS, and startup data reduction codes which include FLOWRATE, COREPERF, CALMET, and VARTAV. These codes were implemented on an HP/Apollo model 9000 series 400 workstation at the YGN 3 site and successfully applied to startup testing and core management. The startup codes made a great contribution in upgrading the reliability of test results and reducing the test period by taking and analyzing core data automatically. The data reduction code saved the manpower and time for test data reduction and decreased the chance for error in the analysis. It is expected that this code system will make similar contributions for reducing the startup testing duration of YGN 4 and UCN3,4

SCORPIO: a framework for core surveillance systems

International Nuclear Information System (INIS)

Berg, Oe.; Tsuiki, Makoto

1999-01-01

The first version of the core surveillance system SCORPIO was installed at Unit 2, Ringhals, in 1984. It was implemented on Norsk Data mini-computers with a fully graphical user-interface. The main purpose was to provide a practical tool for reactor operators and reactor physicists for on-line monitoring and predictive analysis of core behaviour. A second version of SCORPIO was developed in 1993-1995 and implemented on Unix workstations. In addition to upgrading the system at Ringhals, the system was installed by Duke Power, USA, on 7 reactors. SCORPIO was also installed on the Sizewell B reactor. Recently a new framework has been developed which further enhances the flexibility and capabilities for implementing core surveillance systems in different types of nuclear power plants. Modules can be added and replaced in an easy manner. It allows fast and reliable communication of data between modules based on the Software Bus tool developed by IFE. Further, the Picasso-3 user interface management system supports efficient implementation of different user interfaces. Both Unix and Windows NT platforms are supported. The new framework has been applied in development and installation of a SCORPIO-VVER version for the Dukovany NPP, Czech Republic. Here it was of particular importance to provide a flexible system for integration of modules originating from different companies. Development of a BWR version is now in progress. This means that SCORPIO will be available for all the major reactor types, and synergy is obtained by application of a common framework both with respect to system implementation and maintenance. By using the SCORPIO framework, the development time is reduced and the maintenance work is carried out more efficiently, compared to developing systems with lower-level tools. For instance, the MMI can be developed and tested independently of the physics modules

Development of a Web-based CANDU Core Management Procedure Automation System

International Nuclear Information System (INIS)

Lee, Sanghoon; Kim, Eunggon; Park, Daeyou; Yeom, Choongsub; Suh, Hyungbum; Kim, Sungmin

2006-01-01

CANDU reactor core needs efficient core management to increase safety, stability, high performance as well as to decrease operational cost. The most characteristic feature of CANDU is so called 'on-power refueling' i.e., there is no shutdown during refueling in opposition to that of PWR. Although this on-power refueling increases the efficiency of the plant, it requires heavy operational task and difficulties in real time operation such as regulating power distribution, burnup distribution, LZC statistics, the position of control devices and so on. To enhance the CANDU core management, there are several approaches to help operator and reduce difficulties, one of them is the COMOS (CANDU Core On-line Monitoring System). It has developed as an online core surveillance system based on the standard incre instrumentation and the numerical analysis codes such as RFSP (Reactor Fueling Simulation Program). As the procedure is getting more complex and the number of programs is increased, it is required that integrated and cooperative system. So, KHNP and IAE have been developing a new web-based system which can support effective and accurate reactor operational environment called COMPAS that means CANDU cOre Management Procedure Automation System. To ensure development of successful system, several steps of identifying requirements have been performed and Software Requirement Specification (SRS) document was developed. In this paper we emphasis on the how to keep consistency between the requirements and system products by applying requirement traceability methodology

Measuring Performance of Soft Real-Time Tasks on Multi-core Systems

OpenAIRE

Rafiq, Salman

2011-01-01

Multi-core platforms are well established, and they are slowly moving into the area of embedded and real-time systems. Nowadays to take advantage of multi-core systems in terms of throughput, soft real-time applications are run together with general purpose applications under an operating system such as Linux. But due to shared hardware resources in multi-core architectures, it is likely that these applications will interfere and compete with each other. This can cause slower response times f...

Design and development of the helicity injection system in Versatile Experiment Spherical Torus

International Nuclear Information System (INIS)

Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S.

2015-01-01

Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

Design and development of the helicity injection system in Versatile Experiment Spherical Torus

Energy Technology Data Exchange (ETDEWEB)

Park, JongYoon; An, Younghwa; Jung, Bongki; Lee, Jeongwon; Lee, HyunYoung; Chung, Kyoung-Jae; Na, Yong-Su; Hwang, Y.S., E-mail: yhwang@snu.ac.kr

2015-10-15

Graphical abstract: - Highlights: • A high current electron gun with single pulse power for both arc and extraction is developed. • The optimal gun operation is confirmed by impedance matching between the PFN and plasma. • The gun injected currents of 0.95 kA with the voltage of ∼410 V for 5 ms with a 1.2 kV PFN. • The helicity injection system using the gun has been developed and tested successfully in VEST. • Toroidal currents of up to 3.8 kA confirm possible relaxation into tokamak-like plasma. - Abstract: A helicity injection system for the Versatile Experiment Spherical Torus (VEST) has been successfully developed and commissioned. A high current electron gun utilizing hollow cathode and washer stacks has been designed and constructed with a single pulse power system that can provide voltages for both arc discharge and extraction sequentially. Tests for electron gun operation with the single pulse power system have been conducted under various toroidal and poloidal field strengths. The estimated plasma impedance, depending on the injection magnetic field structure, can be utilized for the optimal gun operation by impedance matching between the pulse power system and plasma. With the charging voltage of 1.2 kV, injection current of 0.95 kA has been obtained with the injection voltage of 410 V for about 5 ms. Initial helicity injection experiments have been conducted under various toroidal and poloidal field strengths and a toroidal plasma current of up to 3.8 kA is observed with the current multiplication larger than the geometric stacking ratio, confirming the possibility of relaxation into tokamak-like plasma with closed flux formation.

Steel septum magnets for the LHC beam injection and extraction

CERN Document Server

Bidon, S; Guinand, M; Gyr, Marcel; Sassowsky, M; Weisse, E; Weterings, W; Abramov, A; Ivanenko, A I; Kolatcheva, E; Lapyguina, O; Ludmirsky, E; Mishina, N; Podlesny, P; Riabov, A; Tyurin, N

2002-01-01

The Large Hadron Collider (LHC) will be a superconducting accelerator and collider to be installed in the existing underground LEP ring tunnel at CERN. It will provide proton-proton collisions with a centre of mass energy of 14 TeV. The proton beams coming from the SPS will be injected into the LHC at 450 GeV by vertically deflecting kicker magnets and horizontally deflecting steel septum magnets (MSI). The proton beams will be dumped from the LHC with the help of two extraction systems comprising horizontally deflecting kicker magnets and vertically deflecting steel septum magnets (MSD). The MSI and MSD septa are laminated iron-dominated magnets using an all welded construction. The yokes are constructed from two different half cores, called coil core and septum core. The septum cores comprise circular holes for the circulating beams. This avoids the need for careful alignment of the usually wedge-shaped septum blades used in classical Lambertson magnets. The MSI and MSD septum magnets were designed and buil...

High pressure common rail injection system modeling and control.

Science.gov (United States)

Wang, H P; Zheng, D; Tian, Y

2016-07-01

In this paper modeling and common-rail pressure control of high pressure common rail injection system (HPCRIS) is presented. The proposed mathematical model of high pressure common rail injection system which contains three sub-systems: high pressure pump sub-model, common rail sub-model and injector sub-model is a relative complicated nonlinear system. The mathematical model is validated by the software Matlab and a virtual detailed simulation environment. For the considered HPCRIS, an effective model free controller which is called Extended State Observer - based intelligent Proportional Integral (ESO-based iPI) controller is designed. And this proposed method is composed mainly of the referred ESO observer, and a time delay estimation based iPI controller. Finally, to demonstrate the performances of the proposed controller, the proposed ESO-based iPI controller is compared with a conventional PID controller and ADRC. Copyright © 2016 ISA. Published by Elsevier Ltd. All rights reserved.

Baseline Design Compliance Matrix for the Rotary Mode Core Sampling System

International Nuclear Information System (INIS)

LECHELT, J.A.

2000-01-01

The purpose of the design compliance matrix (DCM) is to provide a single-source document of all design requirements associated with the fifteen subsystems that make up the rotary mode core sampling (RMCS) system. It is intended to be the baseline requirement document for the RMCS system and to be used in governing all future design and design verification activities associated with it. This document is the DCM for the RMCS system used on Hanford single-shell radioactive waste storage tanks. This includes the Exhauster System, Rotary Mode Core Sample Trucks, Universal Sampling System, Diesel Generator System, Distribution Trailer, X-Ray Cart System, Breathing Air Compressor, Nitrogen Supply Trailer, Casks and Cask Truck, Service Trailer, Core Sampling Riser Equipment, Core Sampling Support Trucks, Foot Clamp, Ramps and Platforms and Purged Camera System. Excluded items are tools such as light plants and light stands. Other items such as the breather inlet filter are covered by a different design baseline. In this case, the inlet breather filter is covered by the Tank Farms Design Compliance Matrix

Rates of fuel discharge as affected by the design of fuel-injection systems for internal-combustion engines

Science.gov (United States)

Gelalles, A G; Marsh, E T

1933-01-01

Using the method of weighing fuel collected in a receiver during a definite interval of the injection period, rates of discharge were determined, and the effects noted, when various changes were made in a fuel-injection system. The injection system consisted primarily of a by-pass controlled fuel pump and an automatic injection valve. The variables of the system studied were the pump speed, pump-throttle setting, discharge-orifice diameter, injection-valve opening and closing pressures, and injection-tube length and diameter.

Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

1996-09-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

New system of the in core monitoring - PTK SVRK

International Nuclear Information System (INIS)

Urban, P.

2000-01-01

In this paper author describes new system (PTK SVRK) for in-core monitoring system of the Mochovce nuclear power plant installed instead of the HINDUKUSH in-core monitoring system, which are determined to monitor the core parameters. This system (HINDUKUSH), supplied by the Russian party in scope of the original design, became old during the idle time, and the components, which is it built from, are not produced any more. Thus, its utilisation had to undergo a technical end economic analysis. It resulted in classification to the work complex of the technical specification of safety measures. Its implementation conditioned the commissioning of the power plant nuclear unit. The program and technical system of the in-core monitoring (PTK SVRK) consists of two levels - a 'closed' basic, which fulfils the task of the primal system operation for the Unit operators, and an 'open' top level, which serves as a tool for the additional tasks of a prognosis, monitoring, and analysis of the processes taking place in the nuclear core by the monitoring physicists. The basic level of PTK SVRK has 100% redundancy because of its composition and configuration. It is namely formed by two identical, equivalent, and independent sets. Any of them may be operational or redundant. Every set consists of an apparatus processing the signals coming from the technology or the calculation complex, which converts these signals to physical parameters and controls the physically mathematical model of the monitored equipment. The results are presented to the operational staff as outputs on the workstations in the control room in a form of cartograms, graphs, histograms, tables, etc. The bases of the system calculation model are time-proven programs BIPR7 and PERMAK, which are used also in this power plant. The top level of PTK SVRK has a structure supporting the system openness for its further utilisation. Today it is formed by a server and two workstations. Besides the above-mentioned tasks, the

Strontium-rich injectable hybrid system for bone regeneration

Energy Technology Data Exchange (ETDEWEB)

Neves, Nuno, E-mail: nsmneves@gmail.com [Instituto de Investigação e Inovação em Saúde, Universidade do Porto (Portugal); INEB — Instituto de Engenharia Biomédica, Universidade do Porto, Rua do Campo Alegre 823, 4150-180 Porto (Portugal); FMUP — Faculdade de Medicina da Universidade do Porto, Departamento de Cirurgia, Serviço de Ortopedia, Alameda Prof. Hernâni Monteiro, 4200-319 Porto (Portugal); Campos, Bruno B. [FCUP — Faculdade de Ciências da Universidade do Porto, Centro de Investigação em Química, Departamento de Química e Bioquímica, Rua do Campo Alegre 1021/1055, 4169-007 Porto (Portugal); Almeida, Isabel F.; Costa, Paulo C. [FFUP — Faculdade de Farmácia da Universidade do Porto, Laboratório de Tecnologia Farmacêutica, Departamento de Ciências do Medicamento, Rua de Jorge Viterbo Ferreira 228, 4050-313 Porto (Portugal); Cabral, Abel Trigo [FMUP — Faculdade de Medicina da Universidade do Porto, Departamento de Cirurgia, Serviço de Ortopedia, Alameda Prof. Hernâni Monteiro, 4200-319 Porto (Portugal); and others

2016-02-01

Current challenges in the development of scaffolds for bone regeneration include the engineering of materials that can withstand normal dynamic physiological mechanical stresses exerted on the bone and provide a matrix capable of supporting cell migration and tissue ingrowth. The objective of the present work was to develop and characterize a hybrid polymer–ceramic injectable system that consists of an alginate matrix crosslinked in situ in the presence of strontium (Sr), incorporating a ceramic reinforcement in the form of Sr-rich microspheres. The incorporation of Sr in the microspheres and in the vehicle relies on the growing evidence that Sr has beneficial effects in bone remodeling and in the treatment of osteopenic disorders and osteoporosis. Sr-rich porous hydroxyapatite microspheres with a uniform size and a mean diameter of 555 μm were prepared, and their compression strength and friability tested. A 3.5% (w/v) ultrapure sodium alginate solution was used as the vehicle and its in situ gelation was promoted by the addition of calcium (Ca) or Sr carbonate and Glucone-δ-lactone. Gelation times varied with temperature and crosslinking agent, being slower for Sr than for Ca, but adequate for injection in both cases. Injectability was evaluated using a device employed in vertebroplasty surgical procedures, coupled to a texture analyzer in compression mode. Compositions with 35% w of microspheres presented the best compromise between injectability and compression strength of the system, the force required to extrude it being lower than 100 N. Micro CT analysis revealed a homogeneous distribution of the microspheres inside the vehicle, and a mean inter-microspheres space of 220 μm. DMA results showed that elastic behavior of the hybrid is dominant over the viscous one and that the higher storage modulus was obtained for the 3.5%Alg–35%Sr-HAp-Sr formulation. - Highlights: • We developed a Sr rich viscoelastic hybrid system (alginate matrix crosslinked in

Reactor core flow rate control system

International Nuclear Information System (INIS)

Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

1996-01-01

When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

Design of Electrical System for Inhibitor Injection Pump’s Motor PAQ 01/02/03 RSG-GAS

International Nuclear Information System (INIS)

Taufiq, M.; Teguh Sulistyo; Kiswanto; Santosa Pujiarta

2008-01-01

In order to control the water quality related to the growth of scale, corrosion and micro-organism in the PA01 BR01 and PA02 BR 02 piping system of secondary cooling system of RSG-GAS, electrical systems for motor of inhibitor injection pump PAQ 01/02/03, including motor control system circuit for inhibitor injection pump PAQ02 AP01, motor control system circuit for NaOCl injection pump PAQ01 AP01, motor control system circuit for inhibitor injection pump PAQ02 AP02 and control system circuit for stir pump RW02 have been designed. Motor control system circuit for pump PAQ02 AP01 which attached at the inhibitor tank will operate when conductivity control CQ01 indicates blow down condition and pump motor PAQ02 AP02 is not operate when level control CL02 indicates minimum level. This design is expected that, NaOCl injection pump PAQ01 AP 01 will operate continuously and inhibitor injection pump PAQ02 AP02 will operate automatically. (author)

Improvement of open and semi-open core wall system in tall buildings by closing of the core section in the last story

Science.gov (United States)

Kheyroddin, A.; Abdollahzadeh, D.; Mastali, M.

2014-09-01

Increasing number of tall buildings in urban population caused development of tall building structures. One of the main lateral load resistant systems is core wall system in high-rise buildings. Core wall system has two important behavioral aspects where the first aspect is related to reduce the lateral displacement by the core bending resistance and the second is governed by increasing of the torsional resistance and core warping of buildings. In this study, the effects of closed section core in the last story have been considered on the behavior of models. Regarding this, all analyses were performed by ETABS 9.2.v software (Wilson and Habibullah). Considering (a) drift and rotation of the core over height of buildings, (b) total and warping stress in the core body, (c) shear in beams due to warping stress, (d) effect of closing last story on period of models in various modes, (e) relative displacement between walls in the core system and (f) site effects in far and near field of fault by UBC97 spectra on base shear coefficient showed that the bimoment in open core is negative in the last quarter of building and it is similar to wall-frame structures. Furthermore, analytical results revealed that closed section core in the last story improves behavior of the last quarter of structure height, since closing of core section in the last story does not have significant effect on reducing base shear value in near and far field of active faults.

Optimization of a flow injection analysis system for multiple solvent extraction

International Nuclear Information System (INIS)

Rossi, T.M.; Shelly, D.C.; Warner, I.M.

1982-01-01

The performance of a multistage flow injection analysis solvent extraction system has been optimized. The effect of solvent segmentation devices, extraction coils, and phase separators on performance characteristics is discussed. Theoretical consideration is given to the effects and determination of dispersion and the extraction dynamics within both glass and Teflon extraction coils. The optimized system has a sample recovery similar to an identical manual procedure and a 1.5% relative standard deviation between injections. Sample throughput time is under 5 min. These characteristics represent significant improvements over the performance of the same system before optimization. 6 figures, 2 tables

Development of the RSAC Automation System for Reload Core of WH NPP

International Nuclear Information System (INIS)

Choi, Yu Sun; Bae, Sung Man; Koh, Byung Marn; Hong, Sun Kwan

2006-01-01

The Nuclear Design for Reload Core of Westinghouse Nuclear Power Plant consists of 'Reload Core Model Search', 'Safety Analysis(RSAC)', 'NDR(Nuclear Design Report) and OCAP(Operational Core Analysis Package Generation)' phases. Since scores of calculations for various accidents are required to confirm that the safety analysis assumptions are valid, the Safety Analysis(RSAC) is the most important and time and effort consuming phase of reload core design sequence. The Safety Analysis Automation System supports core designer by the automation of safety analysis calculations in 'Safety Analysis' phase(about 20 calculations). More than 10 kinds of codes, APA(ALPHA/PHOENIX/ANC), APOLLO, VENUS, PHIRE XEFIT, INCORE, etc. are being used for Safety Analysis calculations. Westinghouse code system needs numerous inputs and outputs, so the possibility of human errors could not be ignored during Safety Analysis calculations. To remove these inefficiencies, all input files for Safety Analysis calculations are automatically generated and executed by this Safety Analysis Automation System. All calculation notes are generated and the calculation results are summarized in RSAC (Reload Safety Analysis Checklist) by this system. Therefore, The Safety Analysis Automation System helps the reload core designer to perform safety analysis of the reload core model instantly and correctly

Aging study of boiling water reactor high pressure injection systems

International Nuclear Information System (INIS)

Conley, D.A.; Edson, J.L.; Fineman, C.F.

1995-03-01

The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

Evaluation report on SCTF Core-III Test S3-22

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi; Akimoto, Hajime; Ohnuki, Akira; Abe, Yutaka; Murao, Yoshio; Adachi, Hiromichi.

1991-07-01

Two tests (Tests S3-20 and S3-22) were conducted with JAERI's Slab Core Test Facility (SCTF) Core-III in order to investigate water break-through and core cooling behaviors under the intermittent ECC water delivery from the hot legs to one location in the upper plenum and the alternate ECC water delivery to two locations in the upper plenum during reflooding, respectively. This report presents an analysis on Test S3-22 (the alternate case). Subcooled ECC water was injected alternately just above the upper core support plate above Bundles 7 and 8 and Bundles 3 and 4. The total injection rate from both injection ports was the same as that in SCTF Test S3-20 and Test S3-13. Analyzing the test data together with those of Tests S3-13 and S3-20 the following has been found: (1) Alternate break-through occurred immediately corresponding to the alternate ECC water injection except for one period, during which no break-through was observed. However, there observed a difference in break-through behavior that break-through was strong above the low power region, whereas weak above the high power region. (2) Although its break-through behavior was different, nearly the same core cooling as in the continuous or intermittent ECC water delivery case was observed except for the period around quench. (3) Around quench time, degraded core cooling comparing to the continuous or intermittent ECC water delivery case was observed. That is, quench time at the midplane level of the present test was 35 s later than in the continuous case. This is considered to result from decrease in core water inventory caused by water sealing at the cross-over leg. (J.P.N.)

Evaluation report on SCTF Core-III test S3-06

International Nuclear Information System (INIS)

Iwamura, Takamichi; Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Ohnuki, Akira; Adachi, Hiromichi; Murao, Yoshio; Minato, Akihiko; Sakaki, Isao.

1988-10-01

In order to investigate the effect of radial power distribution on the thermal-hydraulic characteristics during the reflood phase of a PWR-LOCA with a combined injection type ECCS, a core cooling separate effect test S3-06 and a combined injection test S3-16-Phase 2 were performed using the Slab Core Test Facility (SCTF) Core-III. The radial power distributions in these two tests simulated a reference distribution for a PWR with a combined injection type ECCS and a steep distribution for a PWR with a cold leg injection type ECCS, respectively. Under the radial power distribution of a PWR with a combined injection type ECCS, the radial power distribution had little effect on the thermal-hydraulic behavior in the two-phase up-flow region due to the approximately flat power distribution in this region (power ratio = 1.04 ∼ 1.08). The overall fluid behavior in the pressure vessel was also little affected by the radial power distribution. On the other hand, under the steep radial power distribution (peak power ratio = 1.36), the degree of heat transfer enhancement in high power bundles in the two-phase up-flow region was dominated by the bundlewise radial power ratio as in the case of a PWR with a cold leg injection type ECCS. (author)

Supporting system for the core restraint of nuclear reactors

International Nuclear Information System (INIS)

Kaser, A.

1973-01-01

The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

HTR core physics and transient analyses by the Panthermix code system

International Nuclear Information System (INIS)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

2005-01-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

Modelling of reactor control and protection systems in the core simulator program GARLIC

International Nuclear Information System (INIS)

Beraha, D.; Lupas, O.; Ploegert, K.

1984-01-01

For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

Core Flight System (CFS) Integrated Development Environment

Data.gov (United States)

National Aeronautics and Space Administration — The purpose of this project is to create an Integrated Development Environment (IDE) for the Core Flight System (CFS) software to reduce the time it takes to...

Embedded 3D Graphics Core for FPGA-based System-on-Chip Applications

DEFF Research Database (Denmark)

Holten-Lund, Hans Erik

2005-01-01

This paper presents a 3D graphics accelerator core for an FPGA based system, and illustrates how to build a System-on-Chip containing a Xilinx MicroBlaze soft-core CPU and our 3D graphics accelerator core. The system is capable of running uClinux and hardware accelerated 3D graphics applications......, and the video display which periodically reads from memory to display the final rendered graphics. The graphics core uses internal scratch-pad memory to reduce its external bandwidth requirement, this is achieved by implementing a tile-based rendering algorithm. Reduced external bandwidth means that the power...

Development and integration of a 50 Hz pellet injection system for the Experimental Advanced Superconducting Tokamak (EAST)

Energy Technology Data Exchange (ETDEWEB)

Yao, Xingjia [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science and Technology of China, Hefei 230029 (China); Chen, Yue [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Jiansheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Vinyar, Igor; Lukin, Alexander [PELIN, Saint-Petersburg (Russian Federation); Yuan, Xiaoling; Li, Changzheng; Liu, Haiqing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2017-01-15

Highlights: • The design of the pumping system fits the operation requirement well not only theoretically but also experimentally. • The data showed that the averaged pellet injection velocity and propellant gas pressure had a relationship submitting to the power function. • The reliability of the injected pellet was mostly around 90% which is higher than the PI-20 system thanks to the improved pumping system and the new pellet fabrication and acceleration system. - Abstract: A 50 Hz pellet injection system, which is designed for edge-localized mode (ELM) control, has been successfully developed and integrated for the Experimental Advanced Superconducting Tokamak (EAST). Pellet injection is achieved by two separated injection system modules that can be operated independently from 1 to 25 Hz. The nominal injection velocity is 250 m/s with a scatter of ±50 m/s at a repetition rate of 50 Hz. A buffer tank and a two-stage differential pumping system of the pellet injection system was designed to increase hydrogen/deuterium ice quality and eliminate the influence of propellant gas on plasma operation, respectively. The pressure of the buffer tank could be pumped to 1 × 10{sup 2} Pa, and the pressure in the second differential chamber could reach 1 × 10{sup −4} Pa during the experiment. Engineering experiments, which consisted of 50 Hz pellet injection and guiding tube mock-up experiments, were also systematically carried out in a laboratory environment and demonstrated that the pellet injection system can reliably inject pellets at a repetitive frequency of 50 Hz.

Relationship between misonidazole toxicity and core temperature in C3H mice

International Nuclear Information System (INIS)

Gomer, C.J.; Johnson, R.J.

1979-01-01

A single intraperitoneal injection of the radiation sensitizer misonidazole at doses greater than 0.5 mg/g was found to produce a transient hypothermic response in C3H mice. An increase in the acute toxicity of this drug was demonstrated when the animal core temperature was maintained at a normal 35 to 37 0 C by placing the mice in a warmed environment immediately following injection of the drug. The LD/sub 50/3 days/ dose of misonidazole was determined to be 1.48 mg/g for mice allowed to become hypothermic following injection but 0.77 mg/g for mice maintained at a normal core temperature following injection

Development of Core Monitoring System for Nuclear Power Plants (I)

Energy Technology Data Exchange (ETDEWEB)

Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

1997-12-31

1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

A new CF-IRMS system for quantifying stable isotopes of carbon monoxide from ice cores and small air samples

Directory of Open Access Journals (Sweden)

Z. Wang

2010-10-01

Full Text Available We present a new analysis technique for stable isotope ratios (δ¹³C and δ¹⁸O of atmospheric carbon monoxide (CO from ice core samples. The technique is an online cryogenic vacuum extraction followed by continuous-flow isotope ratio mass spectrometry (CF-IRMS; it can also be used with small air samples. The CO extraction system includes two multi-loop cryogenic cleanup traps, a chemical oxidant for oxidation to CO₂, a cryogenic collection trap, a cryofocusing unit, gas chromatography purification, and subsequent injection into a Finnigan Delta Plus IRMS. Analytical precision of 0.2‰ (±1δ for δ¹³C and 0.6‰ (±1δ for δ¹⁸O can be obtained for 100 mL (STP air samples with CO mixing ratios ranging from 60 ppbv to 140 ppbv (~268–625 pmol CO. Six South Pole ice core samples from depths ranging from 133 m to 177 m were processed for CO isotope analysis after wet extraction. To our knowledge, this is the first measurement of stable isotopes of CO in ice core air.

Assessment of MARS for Direct Contact Condensation in the Core Make-up Tank

International Nuclear Information System (INIS)

Park, Keun Tae; Park, Ik Kyu; Lee, Seung Wook

2013-01-01

In order to improve safety features under loss of coolant accident (LOCA) conditions, in many advanced light water reactors, gravity driven passive safety injection systems (PSISs) replace active pump driven emergency core cooling systems. Among various PSISs, the core make-up tank (CMT) with the pressure balancing line (PBL) and the coolant injection line (IL) represents an effective means of providing core cooling. Because the fluid is always sensing the reactor coolant system (RCS) through the PBL connecting the inlet of the CMT to the pressurizer in the case of CP1300 or to the cold legs in the case of AP600/1000, the CMT can provide cold water at any RCS pressure by gravity force. However, after the initiation of LOCAs, if the injection (or isolation) valve is opened, and the steam from the RCS is jetting into the highly subcooled liquid in the CMT and the enhanced interfacial area results in rapid condensation, which in turn, causes a rapid pressure drop in the CMT. As a result, the CMT pressure becomes less than the RCS pressure, and the injection of the CMT can be delayed until the CMT pressure builds up due to greatly reduced condensation in the CMT by the thermal stratification. In order to identify the parameters having significant effects on the gravity-driven injection and the major condensation modes, Lee and No (1998) conducted the separated effect tests of CMT with a small-scale facility. MARS has been developed as a multi-dimensional thermal-hydraulic (TH) system analysis code for the realistic simulation of two-phase TH transients for pressurized water reactor plants. As the backbones for the MARS code, the RELAP5/MOD3.2 and the COB-RA-TF codes were adopted. Recently, Chun et al. (2013) evaluated performance of the SMART passive safety system for SBLOCA using MARS code. However, it is not clarified that MARS can simulate properly the direct contact condensation in the CMT. Thus, in this study, we assess the analysis capability of the MARS code for

Effects of Starch on Properties of Alumina-based Ceramic Cores

Directory of Open Access Journals (Sweden)

LI Fengguang

2016-12-01

Full Text Available In order to improve the poor leachability of alumina-based ceramic cores, different amount of starch was added to the specimens as pore former. Alumina-based ceramic cores were prepared by hot injection technology using corundum powder as base material, paraffin wax and beeswax as plasticizer, silica powder and magnesium oxide powder as mineralizing agent, wherein the parameters of the hot injection process were as follows:temperature of the slurry was 90℃, hot injection pressure was 0.5 MPa and holding time was 25 s. The effects of starch content on the properties of alumina-based ceramic cores were studied and discussed. The results indicate that during sintering period, the loss of starch in the specimens makes porosity of the alumina-based ceramic cores increase. When starch content increases, the room-temperature flexural strength of the ceramic cores reduces and the apparent porosity increases; the volatile solvent increases and the bulk density decreases. After being sintered at 1560℃ for 2.5 h, room-temperature flexural strength of the alumina-based ceramic cores with starch content of 8%(mass fraction is 24.8 MPa, apparent porosity is 47.98% when the volatile solvent is 1.92 g/h and bulk density is 1.88 g/cm3, the complex properties are optimal.

Emission potentials of future diesel fuel injection systems; Emissionspotentiale zukuenftiger Diesel-Einspritzsysteme

Energy Technology Data Exchange (ETDEWEB)

Schommers, J.; Breitbach, H.; Stotz, M.; Schnabel, M. [DaimlerChrysler AG (Germany)

2007-07-01

The historical evolution of the diesel engine correlates strongly with fuel injection system developments. Mercedes-Benz contributed significantly to the recent success of the diesel engine, being one of the first car manufacturers to introduce a modern common rail diesel engine in the Mercedes C220 CDI in 1997. The excellent characteristics of modern diesel engines resulted in a 50% market share in newly registered cars in Germany. These characteristics have to be further improved in the next years to keep the diesel engine attractive. Emissions and at the same time fuel consumption and noise need to be further reduced, while engine power has to go up. For Mercedes-Benz key steps to reach these goals are lower compression ratio, higher boost pressures, higher exhaust gas recirculation rates and better EGR cooling, multiple injection patterns and components with stable application parameters over lifetime. Important requirements for future fuel injection systems are high spray momentum, good stability over lifetime, good robustness of injected quantities for varying injection patterns and a low shot-to-shot variation of injected quantities. The high spray momentum has to be achieved especially for small injections and for part load operating points with low pressures. Therefore, the needle opening and closing velocities are of special importance. With special focus on the above requirements, different injector concepts were hydraulically evaluated. Both concepts in serial production and under development from system suppliers, as well as Mercedes-Benz developed prototype injector concepts were chosen. The concepts analysed are a servo-hydraulically driven injector with control piston, two servo-hydraulically driven injectors without control piston with differently adjusted hydraulics, and a direct driven injector, where the needle is driven directly from an actuator without servo-hydraulic amplification. The hydraulic investigations show an excellent performance of

Optimal Design and Analysis of the Stepped Core for Wireless Power Transfer Systems

Directory of Open Access Journals (Sweden)

Xiu Zhang

2016-01-01

Full Text Available The key of wireless power transfer technology rests on finding the most suitable means to improve the efficiency of the system. The wireless power transfer system applied in implantable medical devices can reduce the patients’ physical and economic burden because it will achieve charging in vitro. For a deep brain stimulator, in this paper, the transmitter coil is designed and optimized. According to the previous research results, the coils with ferrite core can improve the performance of the wireless power transfer system. Compared with the normal ferrite core, the stepped core can produce more uniform magnetic flux density. In this paper, the finite element method (FEM is used to analyze the system. The simulation results indicate that the core loss generated in the optimal stepped ferrite core can reduce about 10% compared with the normal ferrite core, and the efficiency of the wireless power transfer system can be increased significantly.

Core flow control system for field applications; Sistema de controle de core-flow

Energy Technology Data Exchange (ETDEWEB)

Granzotto, Desiree G.; Adachi, Vanessa Y.; Bannwart, Antonio C.; Moura, Luiz F.M. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil); Sassim, Natache S.D.A. [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Centro de Estudo do Petroleo (CEPETRO); Carvalho, Carlos H.M. [PETROBRAS S.A., Rio de Janeiro, RJ (Brazil)

2008-07-01

The significant heavy oil reserves worldwide and the presently high crude oil prices make it essential the development of technologies for heavy oil production and transportation. Heavy oils, with their inherent features of high viscosity (100- 10,000 cP) and density (below 20 deg API) require specific techniques to make it viable their flow in pipes at high flow rates. One of the simplest methods, which do not require use of heat or diluents, is provided by oil-water annular flow (core-flow). Among the still unsolved issues regarding core-flow is the two-phase flow control in order to avoid abrupt increases in the pressure drop due to the possible occurrence of bad water-lubricated points, and thus obtain a safe operation of the line at the lowest possible water-oil ratio. This work presents results of core flow tests which allow designing a control system for the inlet pressure of the line, by actuating on the water flow rate at a fixed oil flow rate. With the circuit model and the specified controller, simulations can be done to assess its performance. The experiments were run at core-flow circuit of LABPETRO-UNICAMP. (author)

On-line core monitoring system based on buckling corrected modified one group model

International Nuclear Information System (INIS)

Freire, Fernando S.

2011-01-01

Nuclear power reactors require core monitoring during plant operation. To provide safe, clean and reliable core continuously evaluate core conditions. Currently, the reactor core monitoring process is carried out by nuclear code systems that together with data from plant instrumentation, such as, thermocouples, ex-core detectors and fixed or moveable In-core detectors, can easily predict and monitor a variety of plant conditions. Typically, the standard nodal methods can be found on the heart of such nuclear monitoring code systems. However, standard nodal methods require large computer running times when compared with standards course-mesh finite difference schemes. Unfortunately, classic finite-difference models require a fine mesh reactor core representation. To override this unlikely model characteristic we can usually use the classic modified one group model to take some account for the main core neutronic behavior. In this model a course-mesh core representation can be easily evaluated with a crude treatment of thermal neutrons leakage. In this work, an improvement made on classic modified one group model based on a buckling thermal correction was used to obtain a fast, accurate and reliable core monitoring system methodology for future applications, providing a powerful tool for core monitoring process. (author)

Experience from development and operation of the core surveillance systems SCORPIO

International Nuclear Information System (INIS)

Berg, O.; Porsmyr, J.; Adlandsvik, K.A.

1994-01-01

The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behaviour. This can first of all improve plant safety as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. These improvements are obtained by better surveillance of core instrumentation and through detailed calculations of core behaviour using on-line simulators. The SCORPIO system has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The system has been in operation at the Ringhals PWR unit 2 in Sweden since the end of 1987 where it runs on Norsk Data mini-computers. Recently, there has been a renewed interest for SCORPIO mainly determined by the utilities' desire to obtain more economical and flexible plant operation. The SCORPIO system has been transferred to Unix based workstations and integrated with the Picasso-2 graphics system. In addition to Ringhals the new system is currently being installed at Nuclear Electri's Sizewell B PWR in UK and Duke Power's Catawba Unit 1,2 and McGuire Unit 1,2, USA. (author). 7 refs, 5 figs

Reactor core cooling device

International Nuclear Information System (INIS)

Kobayashi, Masahiro.

1986-01-01

Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

Analysis of effect of late water injection on RCS repressurization

International Nuclear Information System (INIS)

Tao Jun; Cao Xuewu

2011-01-01

Effect of late water injection on RCS repressurization during high pressure severe accident sequence in a typical PWR was analyzed. As the results shown, late water injection could increase RCS pressure when RPV failed without RCS passive depressurization. Especially in the condition of opening one PORV, RCS pressure could reach high pressure limit when RPV failed and the risk of HPME and DCH was dramatically increased. Integrity of containment could be threatened. However, in the condition of RCS passive depressurization induced by pressurizer surge line creep failure, RCS pressure could be decreased to very low level even only one PORV was opened and two trains of emergency core cooling were implemented. The risk of HPME and DCH was eliminated. The more PORVs were opened, the faster accident progression was and the earlier RPV failed. RCS pressure was a little higher when PRV failed if two trains of emergency core cooling was implemented comparing with the condition with only one train of emergency core cooling. However the time of RPV failure was obviously delayed. From the point of delaying RPV failure and preventing containment early failure of view, the optimized late water injection was opening three PORVs and implementing two trains of emergency core cooling. (authors)

Regional overpower protection system analysis for a DUPIC fuel CANDU core

International Nuclear Information System (INIS)

Jeong, Chang Joon; Choi, Hang Bok; Park, Jee Won

2003-06-01

The regional overpower protection (ROP) system was assessed a CANDU 6 reactor with the DUPIC fuel, including the validation of the WIMS/RFSP/ROVER-F code system used for the estimation of ROP trip setpoint. The validation calculation has shown that it is valid to use the WIMS/RFSP/ROVER-F code system for ROP system analysis of the CANDU 6 core. For the DUPIC core, the ROP trip setpoint was estimated to be 125.7%, which is almost the same as that of the standard natural uranium core. This study has shown that the DUPIC fuel does not hurt the current ROP trip setpoint designed for the natural uranium CANDU 6 reactor

HTR core physics and transient analyses by the Panthermix code system

Energy Technology Data Exchange (ETDEWEB)

Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

Data archiving system implementation in ITER's CODAC Core System

Energy Technology Data Exchange (ETDEWEB)

Castro, R., E-mail: rodrigo.castro@visite.es [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Abadie, L. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Makushok, Y. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Ruiz, M.; Sanz, D. [Instrumentation and Applied Acoustic Research Group, Technical University of Madrid, Madrid (Spain); Vega, J. [CIEMAT Fusion Program, Avda. Complutense 40, Madrid (Spain); Faig, J. [INDRA Sistemas, S.A. Unid. de Sistemas de Control, Dirección de Tecnología Energética, Madrid (Spain); Román-Pérez, G. [Sgenia, C/Chile, 4 Edificio II, Las Rozas, Madrid (Spain); Simrock, S.; Makijarvi, P. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

2015-10-15

Highlights: • Implementation of ITER's data archiving solution. • Integration of the solution into CODAC Core System. • Data archiving structure. • High efficient data transmission into fast plant system controllers. • Fast control and data acquisition in Linux. - Abstract: The aim of this work is to present the implementation of data archiving in ITER's CODAC Core System software. This first approach provides a client side API and server side software allowing the creation of a simplified version of ITERDB data archiving software, and implements all required elements to complete data archiving flow from data acquisition until its persistent storage technology. The client side includes all necessary components that run on devices that acquire or produce data, distributing and streaming to configure remote archiving servers. The server side comprises an archiving service that stores into HDF5 files all received data. The archiving solution aims at storing data coming for the data acquisition system, the conventional control and also processed/simulated data.

Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

International Nuclear Information System (INIS)

Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

1977-04-01

Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

Dry low NOx combustion system with pre-mixed direct-injection secondary fuel nozzle

Science.gov (United States)

Zuo, Baifang; Johnson, Thomas; Ziminsky, Willy; Khan, Abdul

2013-12-17

A combustion system includes a first combustion chamber and a second combustion chamber. The second combustion chamber is positioned downstream of the first combustion chamber. The combustion system also includes a pre-mixed, direct-injection secondary fuel nozzle. The pre-mixed, direct-injection secondary fuel nozzle extends through the first combustion chamber into the second combustion chamber.

Design process of the nanofluid injection mechanism in nuclear power plants

Directory of Open Access Journals (Sweden)

Bang In Choel

2011-01-01

Full Text Available Abstract Nanofluids, which are engineered suspensions of nanoparticles in a solvent such as water, have been found to show enhanced coolant properties such as higher critical heat flux and surface wettability at modest concentrations, which is a useful characteristic in nuclear power plants (NPPs. This study attempted to provide an example of engineering applications in NPPs using nanofluid technology. From these motivations, the conceptual designs of the emergency core cooling systems (ECCSs assisted by nanofluid injection mechanism were proposed after following a design framework to develop complex engineering systems. We focused on the analysis of functional requirements for integrating the conventional ECCSs and nanofluid injection mechanism without loss of performance and reliability. Three candidates of nanofluid-engineered ECCS proposed in previous researches were investigated by applying axiomatic design (AD in the manner of reverse engineering and it enabled to identify the compatibility of functional requirements and potential design vulnerabilities. The methods to enhance such vulnerabilities were referred from TRIZ and concretized for the ECCS of the Korean nuclear power plant. The results show a method to decouple the ECCS designs with the installation of a separate nanofluids injection tank adjacent to the safety injection tanks such that a low pH environment for nanofluids can be maintained at atmospheric pressure which is favorable for their injection in passive manner.

Status of reactor core design code system in COSINE code package

Energy Technology Data Exchange (ETDEWEB)

Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

2014-07-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Status of reactor core design code system in COSINE code package

International Nuclear Information System (INIS)

Chen, Y.; Yu, H.; Liu, Z.

2014-01-01

For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

Combustion characteristics of a gasoline engine with independent intake port injection and direct injection systems for n-butanol and gasoline

International Nuclear Information System (INIS)

He, Bang-Quan; Chen, Xu; Lin, Chang-Lin; Zhao, Hua

2016-01-01

Highlights: • Different injection approaches for n-butanol and gasoline affect combustion events. • High n-butanol percentage in the total energy of fuels improves combustion stability. • N-butanol promotes ignition and shortens combustion duration. • Lean burn increases indicated mean effective pressure at fixed total energy of fuels. • Different fuel injection methods slightly affect indicated mean effective pressure. - Abstract: N-butanol, as a sustainable biofuel, is usually used as a blend with gasoline in spark ignition engines. In this study, the combustion characteristics were investigated on a four-cylinder spark ignition gasoline engine with independent port fuel injection and direct injection systems for n-butanol and gasoline in different operating conditions. The results show that in the case of port fuel injection of n-butanol with direct injection gasoline at a given total energy released in a cycle, indicated mean effective pressure is slightly affected by spark timing at stoichiometry while it changes much more with delayed spark timing in lean burn conditions and is much higher in lean burn conditions compared to stoichiometry at given spark timings. With the increase of n-butanol percentage in a fixed total energy released in a cycle at given spark timings, ignition timing advances, combustion duration shortens, indicated mean effective pressure and indicated thermal efficiency increase. For the cases of port fuel injection of n-butanol with direction injection gasoline and port fuel injection of gasoline with direction injection n-butanol at a fixed total energy released in a cycle, their indicated mean effective pressures are close. But their combustion processes are dependent on fuel injection approaches.

Development of on-line core performance evaluation system for 'FUGEN'

International Nuclear Information System (INIS)

Natori, Hisahide; Kaneto, Kunikazu; Oteru, Shigeru.

1982-01-01

An on-line core performance evaluation system ATROPOS has been developed in order to carry out safe and efficient reactor operation of ''FUGEN''(a heavy water moderated, boiling light water cooled, pressure tube type reactor). This system offers detailed and useful information on such items of core performance as core thermal power, power distribution and thermal operation limits. The power distribution is calculated first by using a three-dimensional nodal coupling model, employing such process data as control rod position and 10 B concentration in the D 2 O moderator. Then the calculated power distribution is corrected by local power monitor readings. An axial one-dimensional nodal coupling model, which considers radial power distribution, and a localized three-dimensional nodal coupling model are used to predict the core thermal power and the power distribution for the region surrounding the control rods respectively, within a short time in advance of control rod operation. The methods employed in this system are verified by comparison with start-up test data from the FUGEN initial core. The estimated power distribution and channel flow agree with values measured by the power calibration monitor and with channel flow converted from measured values of pressure drop, within 3 and 5% respectively, in their root mean square values. The difference in core thermal power between the predicted value and the value measured by the total power monitor is about 1% for control rod operation. (author)

Computer-controlled data acquisition system for the ISX-B neutral injection system

International Nuclear Information System (INIS)

Edmonds, P.H.; Sherrill, B.; Pearce, J.W.

1980-05-01

A data acquisition system for the Impurity Study Experiment (ISX-B) neutral injection system at the Oak Ridge National Laboratory is presented. The system is based on CAMAC standards and is controlled by a MIK-11/2 microcomputer. The system operates at the ion source high voltage on the source table, transmitting the analyzed data to a terminal at ground potential. This reduces the complexity of the communications link and also allows much flexibility in the diagnostics and eventual control of the beam line

EXPERIMENTAL TARGET INJECTION AND TRACKING SYSTEM CONSTRUCTION AND SINGLE SHOT TESTING

International Nuclear Information System (INIS)

PETZOLDT, R.W.; ALEXANDER, N.B.; DRAKE, T.J.; GOODIN, D.T; JONESTRACK, K; VERMILLION, B.A

2003-01-01

Targets must be injected into an IFE power plant at a rate of approximately 5 to 10 Hz. Targets must be tracked very accurately to allow driver beams to be aligned with defined points on the targets with accuracy ± 150 (micro)m for indirect drive and ± 20 (micro)m for direct drive. An experimental target injection and tracking system has been constructed at General Atomics. The injector system will be used as a tool for testing the survivability of various target designs and provide feedback to the target designers. Helium gas propels the targets down an 8 m gun barrel up to 400 m/s. Direct-drive targets are protected in the barrel by sabots that are spring loaded to separate into two halves after acceleration. A sabot deflector directs the sabot halves away from the target injection path. Targets will be optically tracked with laser beams and line-scan cameras. Target position and arrival time will be predicted in real time based on early target position measurements. The system installation will be described. System testing to overcome excessive projectile wear and debris in the gun barrel is presented

Test-Access Planning and Test Scheduling for Embedded Core-Based System Chips

OpenAIRE

Goel, Sandeep Kumar

2005-01-01

Advances in the semiconductor process technology enable the creation of a complete system on one single die, the so-called system chip or SOC. To reduce time-to-market for large SOCs, reuse of pre-designed and pre-veried blocks called cores is employed. Like the design style, testing of SOCs can be best approached in a core-based fashion. In order to enable core-based test development, an embedded core should be isolated from its surrounding circuitry and electrical test access from chip pins...

Development of pellet injection systems for ITER

International Nuclear Information System (INIS)

Combs, S.K.; Gouge, M.J.; Baylor, L.R.

1995-01-01

Oak Ridge National Laboratory (ORNL) has been developing innovative pellet injection systems for plasma fueling experiments on magnetic fusion confinement devices for about 20 years. Recently, the ORNL development has focused on meeting the complex fueling needs of the International Thermonuclear Experimental Reactor (ITER). In this paper, we describe the ongoing research and development activities that will lead to a ITER prototype pellet injector test stand. The present effort addresses three main areas: (1) an improved pellet feed and delivery system for centrifuge injectors, (2) a long-pulse (up to steady-state) hydrogen extruder system, and (3) tritium extruder technology. The final prototype system must be fully tritium compatible and will be used to demonstrate the operating parameters and the reliability required for the ITER fueling application

Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

International Nuclear Information System (INIS)

Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

2006-01-01

The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

Fuel injection system for internal combustion engines. Kraftstoffeinspritzsystem fuer Brennkraftmaschinen

Energy Technology Data Exchange (ETDEWEB)

Hafner, U.

1990-09-13

A fuel injection system for an internal combustion engine is provided with a fuel supply line (13) and at least one electromagnetically actuated fuel injection valve (14) for apportioning a quantity of fuel for injection. A connection muzzle (24) coming from the valve body (23) juts into an opening (22) in the suction pipe (21) of the internal combustion engine. The end of the injection valve opposite the connecting muzzle (24) is connected with the fuel supply line via a fuel entry. The valve body (23) is enclosed by a casing (25) in order to provide the conditions required for a warm start. An annulus (31) extending over a large part of the axial length of the valve remains between the casing and the valve body (23). The annulus (31) communicates with the fuel flow through the fuel supply line (13) via an afflux and an efflux opening (32, 33) (Fig. 1).

Comparative Investigation on 0.4 inch SBLOCA Scenario with Single and Dual Train Passive Safety Injection Systems using SMART-ITL

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yun, Eun-Koo; Choi, Nam-Hyun; Min, Kyoung-Ho; Shin, Yong-Cheol; Bang, Yoon-Gon; Kim, Myoung-Jun; Seo, Chan-Jong; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The Standard Design Approval (SDA) for SMART was certificated in 2012 at the Korea Atomic Energy Research Institute (KAERI). In December 2015, Saudi Arabia and Korea started conducting a three-year project of Pre-Project Engineering (PPE) to prepare a Preliminary Safety Analysis Report (PSAR) and to review the feasibility of constructing SMART reactors in Saudi Arabia. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMART-ITL. In this paper, a comparative investigation was performed on 0.4 inch SBLOCA scenario with single and dual train passive safety injection systems using SMART-ITL. In this paper, the effect of the train number of PSIS on a SBLOCA scenario is investigated for a break size of 0.4 inch. The single and dual train tests show a similar trend in general but the injected water migrates slightly differently in the RV and is discharged through the break nozzle. The parameters of the RV pressure, RV water level, accumulated break mass, and injection flowrates from the CMT and SIT were compared. Compared with the single train test, the increased injection rates from the two trains of the PSIS during the dual train test raised the RV water level, ensuring the safety of the reactor core.

The Injection System of the INFN-SuperB Factory Project: Preliminary Design

Energy Technology Data Exchange (ETDEWEB)

Boni, Roberto; /INFN, Rome; Guiducci, Susanna; /INFN, Rome; Preger, Miro; /INFN, Rome; Raimondi, Pantaleo; /INFN, Rome; Chance, Antoine; /Saclay; Dadoun, Olivier; /Orsay, LAL; Poirier, Freddy; /Orsay, LAL; Variola, Alessandro; /Orsay, LAL; Seeman, John; /SLAC

2012-07-05

The ultra high luminosity B-factory (SuperB) project of INFN requires a high performance and reliable injection system, providing electrons at 4 GeV and positrons at 7 GeV, to fulfil the very tight requirements of the collider. Due to the short beam lifetime, continuous injection of electron and positron bunches in both LER and HER rings is necessary to maintain an high average luminosity. Polarized electrons are required for experiments and must be delivered by the injection system, due to the beam lifetime shorter than the ring polarization build-up: they will be produced by means of a SLAC-SLC polarized gun. The emittance and the energy spread of the e{sup -}/e{sup +} beams are reduced in a 1 GeV Damping Ring (DR) before injection in the main rings. Two schemes for positron production are under study, one with e{sup -}/e{sup +} conversion at low energy (< 1 Gev) and one with conversion at 6 GeV and a recirculation line to bring the positrons back to the DR. Acceleration through the Linac is provided by a 2856 MHz RF system made of travelling wave (TW), room temperature accelerating structures.

Pneumatic hydrogen pellet injection system for the ISX tokamak

International Nuclear Information System (INIS)

Milora, S.L.; Foster, C.A.

1979-01-01

We describe the design and operation of the solid hydrogen pellet injection system used in plasma refueling experiments on the ISX tokamak. The gun-type injector operates on the principle of gas dynamic acceleration of cold pellets confined laterally in a tube. The device is cooled by flowing liquid helium refrigerant, and pellets are formed in situ. Room temperature helium gas at moderate pressure is used as the propellant. The prototype device injected single hydrogen pellets into the tokamak discharge at a nominal 330 m/s. The tokamak plasma fuel content was observed to increase by (0.5--1.2) x 10 19 particles subsequent to pellet injection. A simple modification to the existing design has extended the performance to 1000 m/s. At higher propellant operating pressures (28 bars), the muzzle velocity is 20% less than predicted by an idealized constant area expansion process

Preliminary assessment of adjuster system performance in CANDU-6 RUFIC core

International Nuclear Information System (INIS)

Kim, Soon Young; Suk, Ho Chun

2002-07-01

Four operational transients in CANDU-6 RUFIC core have been simulated to assess the adjuster system performance. These transients included startup after a short shutdown, startup after a poison-out shutdown, shim mode operation, and a stepback to 60% full power. Also, an alternative adjuster-banking scheme has been assessed in this report. The alternative adjuster-banking scheme involves rods in Bank 1 and Bank 7 being re-distributed within the two banks. In the alternative adjuster-banking scheme, Bank 1 becomes the heaviest one. The results of the preliminary assessment indicated that the adjuster system as currently designed and installed in the NU core will adequately meet the functional requirements in the RUFIC core. Comparing to the adjuster system performance in the NU core, the total worth of the adjuster in the RUFIC core is reduced, leading to less xenon override capability and shimming capability. However, the overall performance is expected to still be satisfactory. The overall results from the transient studied indicated that the alternative banking scheme does show some better performance characteristics and merits further detailed studies

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

The control system for SSRF injection and extraction

International Nuclear Information System (INIS)

Yuan Qibing; Gu Ming; Wang Ruiping; Cheng Zhihao; Fan Xuerong; Zhu Haijun

2007-01-01

This paper introduces the injection and extraction control system design for SSRF, which is a distributed control system aimed at stability and reliability of the pulse power supplies, PPS (Personnel Protection System) and MPS (Machine Protection System). The hardware environment is mainly based on PLC (Programmable Logic Controller), and ARM (Advanced RISC Machine) is also applied for studying stability of the power supplies. WinCC and EPICS (Experimental Physics and Industrial Control System) have been selected as the platforms of SCADA (Supervisory Control and Data Acquisition). For unifying the interfacing to the control computer, all front-end equipments are connected via Industrial Ethernet. (authors)

Contamination Control of Freeze Shoe Coring System for Collection of Aquifer Sands

Science.gov (United States)

Homola, K.; van Geen, A.; Spivack, A. J.; Grzybowski, B.; Schlottenmier, D.

2017-12-01

We have developed and tested an original device, the freeze-shoe coring system, designed to recover undisturbed samples of water contained in sand-dominated aquifers. Aquifer sands are notoriously difficult to collect together with porewater from coincident depths, as high hydraulic permeability leads to water drainage and mixing during retrieval. Two existing corer designs were reconfigured to incorporate the freeze-shoe system; a Hydraulic Piston (HPC) and a Rotary (RC) Corer. Once deployed, liquid CO­2 contained in an interior tank is channeled to coils at the core head where it changes phase, rapidly cooling the deepest portion of the core. The resulting frozen core material impedes water loss during recovery. We conducted contamination tests to examine the integrity of cores retrieved during a March 2017 yard test deployment. Perfluorocarbon tracer (PFC) was added to the drill fluid and recovered cores were subsampled to capture the distribution of PFC throughout the core length and interior. Samples were collected from two HPC and one RC core and analyzed for PFC concentrations. The lowest porewater contamination, around 0.01% invasive fluid, occurs in the center of both HPC cores. The greatest contamination (up to 10%) occurs at the disturbed edges where core material contacts drill fluid. There was lower contamination in the core interior than top, bottom, and edges, as well as significantly lower contamination in HPC cores that those recovered with the RC. These results confirm that the freeze-shoe system, proposed for field test deployments in West Bengal, India, can successfully collect intact porewater and sediment material with minimal if any contamination from drill fluid.

Secondary air injection system and method

Science.gov (United States)

Wu, Ko-Jen; Walter, Darrell J.

2014-08-19

According to one embodiment of the invention, a secondary air injection system includes a first conduit in fluid communication with at least one first exhaust passage of the internal combustion engine and a second conduit in fluid communication with at least one second exhaust passage of the internal combustion engine, wherein the at least one first and second exhaust passages are in fluid communication with a turbocharger. The system also includes an air supply in fluid communication with the first and second conduits and a flow control device that controls fluid communication between the air supply and the first conduit and the second conduit and thereby controls fluid communication to the first and second exhaust passages of the internal combustion engine.

Evaluation of High Pressure Components of Fuel Injection Systems Using Speckle Interferometry

OpenAIRE

Basara, Adis

2007-01-01

The modern high pressure fuel injection systems installed in engines provide a highly efficient combustion process accompanied by low emissions of exhaust gases and an impressive level of dynamic response. The design and development of mechanical components for such systems pose a great challenge, since they have to operate under extremely high fluctuating pressures (e.g. up to 2000 bar) for a long lifetime (more than 1000 injections per minute). The permanent change between a higher and a lo...

Evaluation report on CCTF Core-II reflood test C2-16 (Run 76)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Hojo, Tsuneyuki; Murao, Yoshio; Sugimoto, Jun.

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2-16 (Run 76), which was conducted on October 23, 1984, with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood. The objectives of the test are to investigate the reflood phenomena with single failure UPI condition and to investigate the effect of the asymmetry of UPI on the reflood phenomena. The test was performed with an asymmetric UPI condition at the injection rate simulating single failure of LPCI pumps. It was observed that, (1) a UPI test simulating no LPCI pump failure gave the slightly lower peak clad temperature than a UPI test simulating single LPCI pump failure, indicating that single LPCI pump failure assumption is conserrative for UPI condition, and (2) an asymmetric UPI lead to a higher core water accumulation and then a higher heat transfer coefficient, resultantly a lower peak clad temperature than a symmetric UPI, indicating that asymmetric UPI does not lead to a poorer core cooling than symmetric UPI. (author)

Adaptation of a load-inject valve for a flow injection chemiluminescence system enabling dual-reagent injection enhances understanding of environmental Fenton chemistry

International Nuclear Information System (INIS)

Jones, Matthew R.; Nightingale, Philp D.; Turner, Suzanne M.; Liss, Peter S.

2013-01-01

Graphical abstract: -- Highlights: •Measurement of multiple components of Fenton chemistry; Fe(II) and H 2 O 2 . •Rapid, quasi-simultaneous analysis enables calculation of environmental kinetics. •Low, nano to pico-molar detection limits with dual analyte analysis. •Able to measure complex matrix samples – organically enriched seawater. •Low cost system with appreciable sensitivity compared to single analyte analysis. -- Abstract: Environmental Fenton chemistry has been poorly constrained within the marine environment at a multi-component level. A simple, unique, reconfiguration of a flow-injection analytical system combined with luminol chemiluminescence allows quasi-simultaneously the measurement, using a single load-inject valve and a single photon multiplier tube, of reduced iron, Fe(II), and hydrogen peroxide. The system enables rapid, every 22 s, measurements with good accuracy at environmentally relevant concentrations, less than 5% relative standard deviations on both a 5 nM Fe(II) standard and a 60 nM hydrogen peroxide standard. Limits of detection were as low as 40 pM Fe(II) and 100 pM hydrogen peroxide. The system showed excellent capability by measuring from within an organic rich seawater the photochemically induced production of Fe(II) and hydrogen peroxide and their subsequent cycling and Fenton like interactions

Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

Energy Technology Data Exchange (ETDEWEB)

Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

1998-02-01

A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

A Novel Technique of Supra Superficial Musculoaponeurotic System Hyaluronic Acid Injection for Lower Face Lifting

OpenAIRE

Sahawatwong, Sinijchaya; Sirithanabadeekul, Punyaphat; Patanajareet, Vasiyapha; Wattanakrai, Penpun; Thanasarnaksorn, Wilai

2016-01-01

Background: Various methods attempting to correct sagging of the lower face focus mainly on manipulation of the superficial musculoaponeurotic System. Each technique has its own limitation. The authors propose a relatively simple, conservative method utilizing hyaluronic acid injection just above the superficial musculoaponeurotic System. Objective: To address a novel hyaluronic injection technique to lift the lower face. Methods: Details of the injection techniques are described. The Positio...

Physical Watermarking for Securing Cyber-Physical Systems via Packet Drop Injections

Energy Technology Data Exchange (ETDEWEB)

Ozel, Omur [Carnegie Mellon Univ., Pittsburgh, PA (United States); Weekrakkody, Sean [Carnegie Mellon Univ., Pittsburgh, PA (United States); Sinopoli, Bruno [Carnegie Mellon Univ., Pittsburgh, PA (United States)

2017-10-23

Physical watermarking is a well known solution for detecting integrity attacks on Cyber-Physical Systems (CPSs) such as the smart grid. Here, a random control input is injected into the system in order to authenticate physical dynamics and sensors which may have been corrupted by adversaries. Packet drops may naturally occur in a CPS due to network imperfections. To our knowledge, previous work has not considered the role of packet drops in detecting integrity attacks. In this paper, we investigate the merit of injecting Bernoulli packet drops into the control inputs sent to actuators as a new physical watermarking scheme. With the classical linear quadratic objective function and an independent and identically distributed packet drop injection sequence, we study the effect of packet drops on meeting security and control objectives. Our results indicate that the packet drops could act as a potential physical watermark for attack detection in CPSs.

Development of the Learning Health System Researcher Core Competencies.

Science.gov (United States)

Forrest, Christopher B; Chesley, Francis D; Tregear, Michelle L; Mistry, Kamila B

2017-08-04

To develop core competencies for learning health system (LHS) researchers to guide the development of training programs. Data were obtained from literature review, expert interviews, a modified Delphi process, and consensus development meetings. The competencies were developed from August to December 2016 using qualitative methods. The literature review formed the basis for the initial draft of a competency domain framework. Key informant semi-structured interviews, a modified Delphi survey, and three expert panel (n = 19 members) consensus development meetings produced the final set of competencies. The iterative development process yielded seven competency domains: (1) systems science; (2) research questions and standards of scientific evidence; (3) research methods; (4) informatics; (5) ethics of research and implementation in health systems; (6) improvement and implementation science; and (7) engagement, leadership, and research management. A total of 33 core competencies were prioritized across these seven domains. The real-world milieu of LHS research, the embeddedness of the researcher within the health system, and engagement of stakeholders are distinguishing characteristics of this emerging field. The LHS researcher core competencies can be used to guide the development of learning objectives, evaluation methods, and curricula for training programs. © Health Research and Educational Trust.

The Faculty of Language Integrates the Two Core Systems of Number.

Science.gov (United States)

Hiraiwa, Ken

2017-01-01

Only humans possess the faculty of language that allows an infinite array of hierarchically structured expressions (Hauser et al., 2002; Berwick and Chomsky, 2015). Similarly, humans have a capacity for infinite natural numbers, while all other species seem to lack such a capacity (Gelman and Gallistel, 1978; Dehaene, 1997). Thus, the origin of this numerical capacity and its relation to language have been of much interdisciplinary interest in developmental and behavioral psychology, cognitive neuroscience, and linguistics (Dehaene, 1997; Hauser et al., 2002; Pica et al., 2004). Hauser et al. (2002) and Chomsky (2008) hypothesize that a recursive generative operation that is central to the computational system of language (called Merge ) can give rise to the successor function in a set-theoretic fashion, from which capacities for discretely infinite natural numbers may be derived. However, a careful look at two domains in language, grammatical number and numerals, reveals no trace of the successor function. Following behavioral and neuropsychological evidence that there are two core systems of number cognition innately available, a core system of representation of large, approximate numerical magnitudes and a core system of precise representation of distinct small numbers (Feigenson et al., 2004), I argue that grammatical number reflects the core system of precise representation of distinct small numbers alone. In contrast, numeral systems arise from integrating the pre-existing two core systems of number and the human language faculty. To the extent that my arguments are correct, linguistic representations of number, grammatical number, and numerals do not incorporate anything like the successor function.

Aqueous Boric acid injection facility of PWR type reactor

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi; Iwami, Masao.

1996-01-01

If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

The Information Systems Core: A Study from the Perspective of IS Core Curricula in the U.S.

Science.gov (United States)

Hwang, Drew; Ma, Zhongming; Wang, Ming

2015-01-01

To keep up with technology changes and industry trends, it is essential for Information Systems (IS) programs to maintain up to date curricula. In doing so, IS educators need to determine what the IS core is and implement it in their curriculum. This study performed a descriptive analysis of 2,229 core courses offered by 394 undergraduate IS…

Improvements to the sodium supply system of a nuclear reactor core

International Nuclear Information System (INIS)

Chevallier, Rene; Marchais, Christian.

1981-01-01

This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

Recommended HPI [High Pressure Injection] rates for the TMI-2 analysis exercise (0 to 300 minutes)

International Nuclear Information System (INIS)

Anderson, J.L.

1987-09-01

An international analysis exercise has been organized to evaluate the ability of nuclear reactor severe accident computer codes to predict the TMI-2 accident sequence and core damage progression during the first 300 minutes of the accident. A required boundary condition for the analysis exercise is the High Pressure Injection or make-up rates into the primary system during the accident. Recommended injection rates for the first 300 minutes of the accident are presented. Recommendations for several sensitivity studies are also presented. 6 refs., 5 figs., 1 tab

Comparison of injection pain caused by the DentalVibe Injection System versus a traditional syringe for inferior alveolar nerve block anaesthesia in paediatric patients.

Science.gov (United States)

Elbay, M; Şermet Elbay, Ü; Yıldırım, S; Uğurluel, C; Kaya, C; Baydemir, C

2015-06-01

To compare paediatric patients' pain during needle insertion and injection in inferior alveoler nerve block (IANB) anaesthesia injected by either a traditional syringe (TS) or the DentalVibe Injection Comfort System (DV). the study was a randomised controlled crossover clinical trial, comprised of 60 children aged 6-12 requiring an operative procedure with IANB anaesthesia on their mandibular molars bilaterally. One of the molar teeth was treated with TS and the contralateral tooth was treated with DV. On each visit, subjective and objective pain was evaluated using the Wond-Baker Faces Pain Rating Scale (PRS) and the Face, Legg, Cry, Consolability Scale (FLACC Scale). Patients were asked which anaesthesia technique they preferred. Data were analysed using Wilcoxon signed rank, Spearman correlation, and Mann-Whitney U tests. There were no statistically significant differences for pain evalution during needle insertion and injection of each injection system. However, a negative correlation was found on the FLACC between age and pain scores during injection after using DV. Paediatric patients experienced similar pain during IANB anaesthesia administered with TS and DV. With increased age, pain values reduced during anaesthetic agent injection with DV according to FLACC. The traditional procedure was preferred to DV in paediatric patients.

Stripping foils for the PSB H- injection system

CERN Document Server

Aiba, M; Goddard, B; Weterings, W

2009-01-01

Beam physics considerations for the stripping foil of the PSB H- injection system are described, including the arguments for the foil type, thickness, geometry and positioning. The foil performance considerations are described, including expected stripping efficiency, emittance growth, energy straggling, temperature and lifetime. The required movement ranges and tolerances are detailed, together with the assumptions used.

Modelling guidelines for core exit temperature simulations with system codes

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

2015-05-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

Methodology for surge pressure evaluation in a water injection system

Energy Technology Data Exchange (ETDEWEB)

Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

2009-07-01

Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

Modeling of reflood of severely damaged reactor core

International Nuclear Information System (INIS)

Bachrata, A.

2012-01-01

The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

Numerical simulation of boron injection in a BWR

Energy Technology Data Exchange (ETDEWEB)

Tinoco, Hernan, E-mail: htb@forsmark.vattenfall.s [Forsmarks Kraftgrupp AB, SE-742 03 Osthammar (Sweden); Buchwald, Przemyslaw [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden); Frid, Wiktor, E-mail: wiktor@reactor.sci.kth.s [Reactor Technology, Royal Institute of Technology, SE-100 44 Stockholm (Sweden)

2010-02-15

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of

Numerical simulation of boron injection in a BWR

International Nuclear Information System (INIS)

Tinoco, Hernan; Buchwald, Przemyslaw; Frid, Wiktor

2010-01-01

The present study constitutes a first step to understand the process of boron injection, transport and mixing in a BWR. It consists of transient CFD simulations of boron injection in a model of the downcomer of Forsmark's Unit 3 containing about 6 million elements. The two cases studied are unintentional start of boron injection under normal operation and loss of offsite power with partial ATWS leaving 10% of the core power uncontrolled. The flow conditions of the second case are defined by means of an analysis with RELAP5, assuming boron injection start directly after the first ECCS injection. Recent publications show that meaningful conservative results may be obtained for boron or thermal mixing in PWRs with grids as coarse as that utilized here, provided that higher order discretization schemes are used to minimize numerical diffusion. The obtained results indicate an apparently strong influence of the scenario in the behavior of the injection process. The normal operation simulation shows that virtually all boron solution flows down to the Main Recirculation Pump inlet located directly below the boron inlet nozzle. The loss of offsite power simulation shows initially a spread of the boron solution over the entire sectional area of the lower part of the downcomer filled with colder water. This remaining effect of the ECCS injection lasts until all this water has left the downcomer. Above this region, the boron injection jet develops in a vertical streak, eventually resembling the injection of the normal operation scenario. Due to the initial spread, this boron injection will probably cause larger temporal and spatial concentration variations in the core. In both cases, these variations may cause reactivity transients and fuel damage due to local power escalation. To settle this issue, an analysis using an extended model containing the downcomer, the MRPs and the Lower Plenum will be carried out. Also, the simulation time will be extended to a scale of several

Experience and evaluation of advanced on-line core monitoring system 'BEACON' at IKATA site

International Nuclear Information System (INIS)

Fujitsuka, Nobumichi; Tanouchi, Hideyuki; Imamura, Yasuhiro; Mizobuchil, Daisuke

1997-01-01

Shikoku Electric Power Company installed BEACON core monitoring system into IKATA unit 3 in May 1994. During its first cycle of core operation, various operational data were obtained including data of some anomalous reactor conditions introduced for the test objective of the plant start-up. This paper presents the evaluation of the BEACON system capability based on this experience. The system functions such as core monitoring and anomaly detection, prediction of future reactor conditions and increased efficiency of core management activities are discussed. Our future plan to utilize the system is also presented. (authors)

Experimental investigations of a single cylinder genset engine with common rail fuel injection system

Directory of Open Access Journals (Sweden)

Gupta Paras

2014-01-01

Full Text Available Performance and emissions characteristics of compression ignition (CI engines are strongly dependent on quality of fuel injection. In an attempt to improve engine combustion, engine performance and reduce the exhaust emissions from a single cylinder constant speed genset engine, a common rail direct injection (CRDI fuel injection system was deployed and its injection timings were optimized. Results showed that 34°CA BTDC start of injection (SOI timings result in lowest brake specific fuel consumption (BSFC and smoke opacity. Advanced injection timings showed higher cylinder peak pressure, pressure rise rate, and heat release rate due to relatively longer ignition delay experienced.

Research on the Core Competitive Power Elements Evaluation System of Green Hotel

OpenAIRE

Hui LIANG

2013-01-01

Green hotel is a new type of hospitality industry development model based on the concept of circular economy and sustainable development. This paper makes an analysis and evaluation of the elements of green hotel core competence, on this basis, constructs the Green Hotel core competitive evaluation index system. The construction of the system is conducive to understand the green hotel’s own competitive advantage objectively, and explore ways to enhance its core competitiveness, providing obje...

Sample and injection manifolds used to in-place test of nuclear air-cleaning system

International Nuclear Information System (INIS)

Qiu Dangui; Li Xinzhi; Hou Jianrong; Qiao Taifei; Wu Tao; Zhang Jirong; Han Lihong

2012-01-01

Objective: According to the regulations of nuclear safety rules and related standards, in-place test of the nuclear air-cleaning systems should be carried out before and during operation of the nuclear facilities, which ensure them to be in good condition. In some special conditions, the use of sample and injection manifolds is required to make the test tracer and ventilating duct air fully mixed, so as to get the on-spot typical sample. Methods: This paper introduces the technology and application of the sample and injection manifolds in nuclear air-cleaning system. Results: Multi point injection and multi point sampling technology as an effective experimental method, has been used in a of domestic and international nuclear facilities. Conclusion: The technology solved the problem of uniformly of on-spot injection and sampling,which plays an important role in objectively evaluating the function of nuclear air-cleaning system. (authors)

Short interval measurement of the Thomson scattering system at the pellet injection by using the event triggering system in LHD

International Nuclear Information System (INIS)

Yasuhara, R.; Sakamoto, R.; Motojima, G.; Yamada, I.; Hayashi, H.

2013-01-01

We have demonstrated Thomson scattering measurements of a short interval less than 1 ms by using the event triggering system with a multi-laser configuration. We have tried to measure this system at the pellet injection and obtained electron temperature and density profiles before and just after the pellet injection. Obtained profiles were dramatically changed after pellet injection with shot-by-shot measurements. This measurement technique will contribute understanding the physics of the pellet deposition. (author)

Experiment on performance of upper head injection system with ROSA-II

International Nuclear Information System (INIS)

1976-09-01

Thermo-hydraulic behavior in the primary cooling system of a pressurized water reactor with an upper head injection system (UHI) in a postulated loss-of-coolant accident (LOCA) has been studied with ROSA-II test facility. Simulated UHI and internal structures of the pressure vessel were installed to the facility for the experiment. Nine maximum-sized double-ended break tests and one medium-sized split break test were performed for the cold-leg break condition. The results are as follows: (1) Fluid mixing in the upper head is not perfect. (2) Cold water injection into the steam or two-phase fluid causes violent depressurization due to the condensation. Flow pattern in the primary cooling system is largely influenced by the above two. (auth.)

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

Development of a core follow calculational system for research reactors

International Nuclear Information System (INIS)

Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

1994-01-01

Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

Legal Protection on IP Cores for System-on-Chip Designs

Science.gov (United States)

Kinoshita, Takahiko

The current semiconductor industry has shifted from vertical integrated model to horizontal specialization model in term of integrated circuit manufacturing. In this circumstance, IP cores as solutions for System-on-Chip (SoC) have become increasingly important for semiconductor business. This paper examines to what extent IP cores of SoC effectively can be protected by current intellectual property system including integrated circuit layout design law, patent law, design law, copyright law and unfair competition prevention act.

A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

International Nuclear Information System (INIS)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

1995-01-01

The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

External injection systems applied in modern cyclotrons designed and manufactured in NIIEFA

International Nuclear Information System (INIS)

Bogdanov, P.V.; Vasilchenko, I.N.; Veresov, O.L.; Gavrish, Yu.N.; Grigorenko, S.V.; Zuev, Yu.V.; Kozienko, M.T.; Mudrolyubov, V.G.; Strokach, A.P.; Tsygankov, S.S.

2012-01-01

The main parameters and design features of the external injection systems applied in modern cyclotrons designed and manufactured in NIIEFA and intended for production of radionuclides for medicine are presented. The use of these external injection systems instead of a traditional internal source allows the current of the accelerated beam to be significantly increased and the in-leakage of the working gas to the acceleration chamber to be reduced, which results in reduced beam losses in the process of acceleration and lower equipment activation.

TARA beamline and injection system

International Nuclear Information System (INIS)

Post, R.S.; Brindza, P.; Coleman, J.W.; Torti, R.P.; Blackfield, D.T.; Goodrich, P.

1983-01-01

The TARA beamline for neutral beam injection will permit one to three sources to fire into each plug (60 degree or optional 90 degree injection with respect to the TARA axis) or into each anchor (90 degree injection only). The sources, pre-aimed on their mounting plate at the NB test stand, may be fired into neutralizer ducts or optionally through a magnesium curtain, and the unneutralized fraction is dumped by the TARA fringing field onto a receiver plate. The beamline is housed in a cylindrical tank with the beam axis along the tank diameter at the midplane. The tank will be sorption pumped using LN + T/sub I/ or N/sub B/ and/or e-beam gettering. The beam burial tank contains sed arrays and a thin foil dump which reaches sufficiently high temperatures during the shot to boil out gas between shots

Repetitive laser fusion experiment and operation using a target injection system

International Nuclear Information System (INIS)

Nishimura, Yasuhiko; Komeda, Osamu; Mori, Yoshitaka

2017-01-01

Since 2008, a collaborative research project on laser fusion development based on a high-speed ignition method using repetitive laser has been carried out with several collaborative research institutes. This paper reports the current state of operation of high repetition laser fusion experiments, such as target introduction and control based on a target injection system that allows free falling under 1 Hz, using a high repetition laser driver that has been under research and development, as well as the measurement of targets that freely fall. The HAMA laser driver that enabled high repetition fusion experiments is a titanium sapphire laser using a diode-pumped solid-state laser KURE-I of green light output as a driver pump light source. In order to carry out high repetition laser fusion experiments, the target injection device allows free falling of deuterated polystyrene solid sphere targets of 1 mm in diameter under 1 Hz. The authors integrated the developed laser and injection system, and succeeded first in the world in making the nuclear fusion reaction continuously by hitting the target to be injected with laser, which is essential technology for future laser nuclear fusion reactor. In order to realize repetition laser fusion experiments, stable laser, target synchronization control, and target position measurement technologies are indispensable. (A.O.)

A Common Definition of the System Operators' Core Activities[Electric Power Transmission System Operator

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-02-15

In this report a common definition of the system operator's core activities in the Nordic countries is identified and also a list of non-core activities is introduced. As a starting point the common tasks for system responsibility as identified by Nordel has been used for the work. The term TSO (Transmission System Operator) is employed as a common denominator in the report. It is found out that the TSOs carry out common core activities in the roles as a transmission operator, a system operator and a balance settlement responsible. The core activities for the TSO as a transmission network operator are: Maintain the adequate transmission system in the long run and network development plan on the national as well as on the Nordic level using sophisticated analysis and planning methods and tools. Plan the transmission network on the national as well as on the Nordic level utilising new investments, renewal and maintenance of existing network components so that the network is secure to operate and adequate transmission capacity is guaranteed. Aim at timely network expansions using enhanced information exchange between the Nordic TSOs, and on the national level between the TSO and distribution and regional network operators, large consumers and large producers. Secure the technical compatibility with networks across the border and within a country by establishing connection requirements on the national level and ensuring that the national requirements are compatible across the Nordic power system. The core activities for the TSO as a system operator are: Define common technical requirements for the secure system operation using common planning, operation, connection and data exchange procedures. Secure the system operation with the operational planning for the following year by using information exchange between TSOs enabling the TSOs to make the best possible forecast of the global grid situation in order to assess the flows in their network and the available

The reactor core configuration and important systems related to physics tests of Daya Bay NPP

International Nuclear Information System (INIS)

Tao Shaoping

1995-06-01

A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

Integrated development environment for multi-core systems

Directory of Open Access Journals (Sweden)

Krunić Momčilo V.

2014-01-01

Full Text Available Development of the software application that provides comfortable working environment of embedded software applications was always a difficult task to achieve. To reach this goal it was necessary to integrate all specific tools designed for that purpose. This paper describes Integrated Development Environment (IDE that was developed to meet all specific needs of a software development for the family of multi-core target platforms designed for a digital signal processing in Cirrus Logic Company. Eclipse platform and RCP (Rich Client Platform was used as a basis, because it provides an extensible plug-in system for customizing the development environment. CLIDE (Cirrus Logic Integrated Development Environment represent the epilog of that effort, reliable IDE used for development of embedded applications. Validation of the solution is accomplished thru 2641 J Unit tests that validate most of the CLIDE's functionalities. Developed IDE (CLIDE significantly increases a quality of a software development for multi-core systems and reduces time-to-market, thereby justifying development costs.

Evaluation report on CCTF core-II reflood test C2 - 18 (Run 78)

International Nuclear Information System (INIS)

Iguchi, Tadashi; Akimoto, Hajime; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun; Hojo, Tsuneyuki.

1987-03-01

This report presents the result of the upper plenum injection (UPI) test C2 - 18 (Run 78), which was conducted on November 13, 1984 with the Cylindrical Core Test Facility (CCTF) at Japan Atomic Energy Research Institute (JAERI). The CCTF is a 1/21.4 scale model of a 1,100 MWe PWR with four loop active components to provide information on the system and core thermo-hydrodynamics during reflood phase. The objectives of the test are to investigate the refill behavior with UPI condition and to investigate the reflood behavior with UPI Best-Estimate (BE) condition. The test was performed to simulate refill/reflood behavior with UPI and BE conditions (However, the LPCI flow rate was determined based on single failure of LPCI pumps.). The result of the test showed the followings. (1) Little special phenomena were recognized under UPI and BE conditions in comparison with those under UPI and Evaluation-Model (EM) conditions, although certain special phenoma (i.e. significant fluid oscillation) were recognized under Cold-Leg-Injection (CLI) and BE conditions in comparison with those under CLI and EM conditions. (2) Water inventory in lower plenum increased smoothly due to water injected into both upper plenum and cold leg during refill phase, similarly to that in refill-simulation test with CLI condition. Small difference in refill behavior with UPI condition is the existing of steam condensation in upper plenum, resulting in lower steam binding and higher core cooling during early reflood phase. This indicates the conservatism of UPI against CLI during early reflood phase. (3) The good core-cooling capability was confirmed under UPI and BE conditions. (author)

A High-Resolution Continuous Flow Analysis System for Polar Ice Cores

DEFF Research Database (Denmark)

Dallmayr, Remi; Goto-Azuma, Kumiko; Kjær, Helle Astrid

2016-01-01

of Polar Research (NIPR) in Tokyo. The system allows the continuous analysis of stable water isotopes and electrical conductivity, as well as the collection of discrete samples from both inner and outer parts of the core. This CFA system was designed to have sufficiently high temporal resolution to detect...... signals of abrupt climate change in deep polar ice cores. To test its performance, we used the system to analyze different climate intervals in ice drilled at the NEEM (North Greenland Eemian Ice Drilling) site, Greenland. The quality of our continuous measurement of stable water isotopes has been......In recent decades, the development of continuous flow analysis (CFA) technology for ice core analysis has enabled greater sample throughput and greater depth resolution compared with the classic discrete sampling technique. We developed the first Japanese CFA system at the National Institute...

Fuel management and core design code systems for pressurized water reactor neutronic calculations

International Nuclear Information System (INIS)

Ahnert, C.; Arayones, J.M.

1985-01-01

A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

Compact multipurpose sub-sampling and processing of in-situ cores with press (pressurized core sub-sampling and extrusion system)

Energy Technology Data Exchange (ETDEWEB)

Anders, E.; Muller, W.H. [Technical Univ. of Berlin, Berlin (Germany). Chair of Continuum Mechanics and Material Theory

2008-07-01

Climate change, declining resources and over-consumption result in a need for sustainable resource allocation, habitat conservation and claim for new technologies and prospects for damage-containment. In order to increase knowledge of the environment and to define potential hazards, it is necessary to get an understanding of the deep biosphere. In addition, the benthic conditions of sediment structure and gas hydrates, temperature, pressure and bio-geochemistry must be maintained during the sequences of sampling, retrieval, transfer, storage and downstream analysis. In order to investigate highly instable gas hydrates, which decomposes under pressure and temperature change, a suite of research technologies have been developed by the Technische Universitat Berlin (TUB), Germany. This includes the pressurized core sub-sampling and extrusion system (PRESS) that was developed in the European Union project called HYACE/HYACINTH. The project enabled well-defined sectioning and transfer of drilled pressure-cores obtained by a rotary corer and fugro pressure corer into transportation and investigation chambers. This paper described HYACINTH pressure coring and the HYACINTH core transfer. Autoclave coring tools and HYACINTH core logging, coring tools, and sub-sampling were also discussed. It was concluded that possible future applications include, but were not limited to, research in shales and other tight formations, carbon dioxide sequestration, oil and gas exploration, coalbed methane, and microbiology of the deep biosphere. To meet the corresponding requirements and to incorporate the experiences from previous expeditions, the pressure coring system would need to be redesigned to adapt it to the new applications. 3 refs., 5 figs.

Heavy mineral sorting in downwards injected Palaeocene sandstone, Siri Canyon, Danish North Sea

DEFF Research Database (Denmark)

Kazerouni, Afsoon Moatari; Friis, Henrik; Svendsen, Johan Byskov

2011-01-01

Post-depositional remobilization and injection of sand are often seen in deep-water clastic systems and has been recently recognised as a significant modifier of deep-water sandstone geometry. Large-scale injectite complexes have been interpreted from borehole data in the Palaeocene Siri Canyon...... of depositional structures in deep-water sandstones, the distinction between "in situ" and injected or remobilised sandstones is often ambiguous. Large scale heavy mineral sorting (in 10 m thick units) is observed in several reservoir units in the Siri Canyon and has been interpreted to represent the depositional...... sorting. In this study we describe an example of effective shear-zone sorting of heavy minerals in a thin downward injected sandstone dyke which was encountered in one of the cores in the Cecilie Field, Siri Canyon. Differences in sorting pattern of heavy minerals are suggested as a tool for petrographic...

CO2 Injectivity in Geological Storages: an Overview of Program and Results of the GeoCarbone-Injectivity Project

International Nuclear Information System (INIS)

Lombard, J.M.; Egermann, P.; Azaroual, M.; Pironon, J.; Broseta, D.; Egermann, P.; Munier, G.; Mouronval, G.

2010-01-01

The objective of the GeoCarbone-Injectivity project was to develop a methodology to study the complex phenomena involved in the near well bore region during CO 2 injection. This paper presents an overview of the program and results of the project, and some further necessary developments. The proposed methodology is based on experiments and simulations at the core scale, in order to understand (physical modelling and definition of constitutive laws) and quantify (calibration of simulation tools) the mechanisms involved in injectivity variations: fluid/rock interactions, transport mechanisms, geomechanical effects. These mechanisms and the associated parameters have then to be integrated in the models at the well bore scale. The methodology has been applied for the study of a potential injection of CO 2 in the Dogger geological formation of the Paris Basin, in collaboration with the other ANR GeoCarbone projects. (authors)

On-line core monitoring with CORE MASTER / PRESTO

International Nuclear Information System (INIS)

Lindahl, S.O.; Borresen, S.; Ovrum, S.

1986-01-01

Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

A benchmark for coupled thermohydraulics system/three-dimensional neutron kinetics core models

International Nuclear Information System (INIS)

Kliem, S.

1999-01-01

During the last years 3D neutron kinetics core models have been coupled to advanced thermohydraulics system codes. These coupled codes can be used for the analysis of the whole reactor system. Although the stand-alone versions of the 3D neutron kinetics core models and of the thermohydraulics system codes generally have a good verification and validation basis, there is a need for additional validation work. This especially concerns the interaction between the reactor core and the other components of a nuclear power plant (NPP). In the framework of the international 'Atomic Energy Research' (AER) association on VVER Reactor Physics and Reactor Safety, a benchmark for these code systems was defined. (orig.)

Sensors Based Measurement Techniques of Fuel Injection and Ignition Characteristics of Diesel Sprays in DI Combustion System

Directory of Open Access Journals (Sweden)

S. Rehman

2016-09-01

Full Text Available Innovative sensor based measurement techniques like needle lift sensor, photo (optical sensor and piezoresistive pressure transmitter are introduced and used to measure the injection and combustion characteristics in direct injection combustion system. Present experimental study is carried out in the constant volume combustion chamber to study the ignition, combustion and injection characteristics of the solid cone diesel fuel sprays impinging on the hot surface. Hot surface ignition approach has been used to create variety of advanced combustion systems. In the present study, the hot surface temperatures were varied from 623 K to 723 K. The cylinder air pressures were 20, 30 and 40 bar and fuel injection pressures were 100, 200 and 300 bar. It is found that ignition delay of fuel sprays get reduced with the rise in injection pressure. The ignition characteristics of sprays much less affected at high fuel injection pressures and high surface temperatures. The fuel injection duration reduces with the increase in fuel injection pressures. The rate of heat release becomes high at high injection pressures and it decreases with the increase in injection duration. It is found that duration of burn/combustion decrease with the increase in injection pressure. The use of various sensors is quite effective, reliable and accurate in measuring the various fuel injection and combustion characteristics. The study simulates the effect of fuel injection system parameters on combustion performance in large heavy duty engines.

Lithium pellet injection experiments on the Alcator C-Mod tokamak

Energy Technology Data Exchange (ETDEWEB)

Garnier, Darren Thomas [Univ. of California, Berkeley, CA (United States)

1996-06-01

A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient (ℓ_p ℓ a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li⁺ emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = ∞ and the n = 1 marginal stability limits.

Lithium pellet injection experiments on the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Garnier, D.T.

1996-06-01

A pellet enhanced performance mode, showing significantly reduced core transport, is regularly obtained after the injection of deeply penetrating lithium pellets into Alcator C-Mod discharges. These transient modes, which typically persist about two energy confinement times, are characterized by a steep pressure gradient (ell p ≤ a/5) in the inner third of the plasma, indicating the presence of an internal transport barrier. Inside this barrier, particle and energy diffusivities are greatly reduced, with ion thermal diffusivity dropping to near neoclassical values. Meanwhile, the global energy confinement time shows a 30% improvement over ITER89-P L-mode scaling. The addition of ICRF auxiliary heating shortly after the pellet injection leads to high fusion reactivity with neutron rates enhanced by an order of magnitude over L-mode discharges with similar input powers. A diagnostic system for measuring equilibrium current density profiles of tokamak plasmas, employing high speed lithium pellets, is also presented. Because ions are confined to move along field lines, imaging the Li + emission from the toroidally extended pellet ablation cloud gives the direction of the magnetic field. To convert from temporal to radial measurements, the 3-D trajectory of the pellet is determined using a stereoscopic tracking system. These measurements, along with external magnetic measurements, are used to solve the Grad-Shafranov equation for the magnetic equilibrium of the plasma. This diagnostic is used to determine the current density profile of PEP modes by injection of a second pellet during the period of good confinement. This measurement indicates that a region of reversed magnetic shear exists at the plasma core. This current density profile is consistent with TRANSP calculations for the bootstrap current created by the pressure gradient. MHD stability analysis indicates that these plasmas are near the n = ∞ and the n = 1 marginal stability limits

Fuel injection and mixing systems having piezoelectric elements and methods of using the same

Science.gov (United States)

Mao, Chien-Pei [Clive, IA; Short, John [Norwalk, IA; Klemm, Jim [Des Moines, IA; Abbott, Royce [Des Moines, IA; Overman, Nick [West Des Moines, IA; Pack, Spencer [Urbandale, IA; Winebrenner, Audra [Des Moines, IA

2011-12-13

A fuel injection and mixing system is provided that is suitable for use with various types of fuel reformers. Preferably, the system includes a piezoelectric injector for delivering atomized fuel, a gas swirler, such as a steam swirler and/or an air swirler, a mixing chamber and a flow mixing device. The system utilizes ultrasonic vibrations to achieve fuel atomization. The fuel injection and mixing system can be used with a variety of fuel reformers and fuel cells, such as SOFC fuel cells.

Evaluation report on SCTF Core-III tests S3-7 and S3-8

International Nuclear Information System (INIS)

Okubo, Tsutomu; Iguchi, Tadashi; Iwamura, Takamichi

1990-03-01

It has been said that the Emergency Core Cooling (ECC) water injected into the hot legs flows into the upper plenum and then falls back to the core (i.e. break-through) during reflood phase in a German type Pressurized Water Reactor (GPWR) with the combined-injection-type ECCS, and that the break-through occurs where the water temperature at the tie plate area is lower and subcooled. Based on this information two tests were conducted with the Slab Core Test Facility (SCTF) Core-III in order to investigate the effects of the water temperature distribution at the tie plate area on the break-through and the core cooling. In these tests, the subcooled ECC water was injected just above the Upper Core Support Plate (UCSP) in order to establish the desired water temperature distribution at the tie plate area. In one test (Test S3-7) the ECC water injection above the UCSP was performed above Bundles 3 and 4, and in the other test (Test S3-8) above Bundles 7 and 8 during initial 60 s a and then was changed to above Bundles 3 and 4. The test data were compared with those of Test S3-SH1, in which the injection was performed above Bundles 7 and 8 and the other test conditions were the same as in Tests S3-7 and S3-8. Analyzing these test data, the following has been found: The break-through occurs where the water temperature at the tie plate area is subcooled and the core cooling is enhanced significantly in the break-through region. The break-through location changes, with some time lag, following the change of the water temperature distribution at the tie plate area. Furthermore, the core cooling in the non-break-through regions is almost the same regardless of the location of the break-through. (author)

Multi-core System Architecture for Safety-critical Control Applications

DEFF Research Database (Denmark)

Li, Gang

and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

System Study: High-Pressure Safety Injection 1998-2014

Energy Technology Data Exchange (ETDEWEB)

Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

2015-12-01

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

Comparative efficacy and safety of local and systemic methotrexate injection in cesarean scar pregnancy

Directory of Open Access Journals (Sweden)

Peng P

2015-01-01

Full Text Available Ping Peng,1 Ting Gui,1 Xinyan Liu,1 Weilin Chen,1 Zhenzhen Liu2 1Department of Obstetrics and Gynecology, 2Department of Ultrasonography, Peking Union Medical College Hospital, Peking Union Medical College, Chinese Academy of Medical Sciences, Beijing, People’s Republic of China Objective: To investigate the efficacy of methotrexate (MTX injection in treatment of cesarean scar pregnancy (CSP. Method: A randomized controlled study was performed in 104 CSP patients receiving either local or systemic MTX injection at the Peking Union Medical College Hospital from the year 2008 to 2013. Results: Complete cure was defined as regression of ultrasonographic findings and normalization of serum β-hCG within 60 days. It was regarded as delayed cure if additional dilation and curettage (D&C was needed. The overall cure rate (complete cure plus delayed cure was 69.2% versus 67.3% for local injection versus systemic administration (P>0.05. The median time for serum β-hCG remission and uterine mass disappearance after systemic administration (42 [21–69] days and 40 [20–67] days were significantly lower than those receiving local injection (56 [24–92] days and 53 [23–88] days, with P=0.029 and 0.046, respectively. The mean pretreatment serum β-hCG (human chorionic gonadotropin level and lesion size in cured group (21,941±18,351 mIU/mL and 2.9±1.3 cm, respectively were significantly lower than those in the failed group (37,047±30,864 mIU/mL and 3.6±1.3 with P=0.038 and 0.044, respectively. Conclusion: MTX injection is effective in CSP treatment. Systemic administration shows similar overall cure rate compared to local injection, but requires shorter time for serum β-hCG remission and uterine mass disappearance. Keywords: cesarean scar pregnancy, methotrexate injection, local, systemic

Transjugular liver biopsy : the efficacy of quick-core biopsy needle system

Energy Technology Data Exchange (ETDEWEB)

Jung, Gyoo Sik; Ahn, Byung Kwon; Lee, Sang Ouk; Chang, Hee Kyong; Oh, Kyung Seung; Huh, Jin Do; Joh, Young Duk [Kosin Medical College, Pusan (Korea, Republic of)

1998-02-01

To evaluate the efficacy of the Quick-Core biopsy needle system in performing transjugular liver biopsy. Between December 1995 and June 1997, eight patients underwent transjugular liver biopsy involving use of the Quick-Core biopsy needle system; the conditions involved were coagulopathy (n=4), thrombocytopenia (n=3), and ascites (n=1). Via the right internal jugular vein, the right hepatic vein was selectively catheterized with a 7-F transjugular guiding catheter, and a14-guage stiffening cannula was then inserted through this catheter; to obtain core tissue, a Quick-Core needle was then advanced into the liver parenchyma through the catheter-cannula combination. Eighteen- and 19-guage needles were used in three and five patients, respectively; specimen size, adequacy of the biopsy specimen and histologic diagnosis were determined, and complications were recorded. Biopsy was successful in all patients. The mean length of the specimen was 1.4 cm (1.0 - 1.8 cm), and all were adequate for pathologic examinations ; specific diagnosis was determined in all patients. There were two malignant neoplasms, two cases of veno-occlusive disease, and one case each of cirrhosis, fulminant hepatitis, Banti syndrome and Budd-Chiari syndrome. One patient complained of neck pain after the procedure, but no serious procedural complications were noted. Our preliminary study shows that the Quick-Core biopsy needle system is safe and provides adequate core tissues with high diagnostic yields. (author). 23 refs., 1 tab., 3 figs.